ML20155A102
| ML20155A102 | |
| Person / Time | |
|---|---|
| Issue date: | 06/01/1988 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-T-1673, NUDOCS 8806090248 | |
| Download: ML20155A102 (396) | |
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OR3\inL UNITED STATES k NUCLEAR REGULATORY COMMISSION 1
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS In the Matter of: )
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SUBCOMMITTEE ON THE ADVANCED )
BOILING WATER REACTOR )
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(OPEN SESSION) 9 .
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DATE: June 1, 1988 PAGES: 1 through 284
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1 PUBLIC NOTICE BY THE 2 UNITED STATES NUCLEAR REGULATORY COMMISSION'S 3 ADVISORY COKMITTEE ON REACTOR SAFEGUARDS 4
5 6'
4 7 The contents of this stenographic transcript of the 8 proceedings of the United States Nuclear Regulatory ,
9 Commission's Advisory Committee on Reactor Safeguards (ACRS),
10 as reported herein, is an uncorrected record of the discussions 11 recorded at the meeting held on the above date.
12 No member of the ACRS Staff and no participant at 13 this meeting accepts any responsibility for errors or
)
14 inaccuracies of statement or data contained in this transcript. ,
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r3 1 1 UNITED STATES NUCLEAR REGULATORY COMMISSION 2 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 3
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4 Meeting of the Subcommittee )
on The Advanced Boiling )
5 Water Reactor )
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6 Wednesday, 7 June 1, 1988 8 1717 H Street, N.W.
Washington, D.C. 20555 9
The above-entitled matter came on for hearing, 10 pursuant to notice, at 8:30 a.m.
11 BEFORE: MR. CARLYLE MICHELSON 12 Chairman
() 13 Retired Principal Nuclear Engineer Tennessee Valley Authority, Knoxville, T3nnessee, and 14 Retired Director, Office for Analysis and Evaluation of Operational Data 15 U.S. Nuclear Regulatory Commission Washington, D.C.
16 ACRS MEMBERS PRESENT:
17 DR. WILLIAM KERR 18 Professor of Nuclear Engineering Director, Office of Energy Research 19 University of Michigan Ann Arbor, Michigan 20 DR. FORREST J. REMICK 21 Vice Chairman Associate Vice-President for Research 22 Professor of Nuclear Engineering The Pennsylvania State University 23 University Park, Pennsylvania 24
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4 (2) 2 1 ACRS MEMBERS PRESENT (CONTINUED):
2 DR. PAUL G. SHEWMON Professor, Metallurgical Engineering Department 3 Ohio State University Columbus, Ohio 4
DR. CHESTER P. SIESS S Professor Emeritus of Civil Engineering University of Illinois 6 Urbana, Illinois 7 MR. DAVID A. WAPD Research Manager on Special Assignment 8 E.I. du Pont de Nemours & Company Savannah River Laboratory 9 Aiken, South Carolina 10 MR. CHARLES J. WYLIE Retired Chief Engineer 11 Electrical Division Duke Power Company 12 Charlotte, North Carolina O
l 13 ACRS COGNIZANT STAFF MEMBER:
14 Richard K. Major 15 CONSULTANT: ,
16 Jesse C. Ebersole 17 18 19 20 21 ,
22 23 24
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3 1 PROCEEDINGS 2 MR. MICHELSON: Good morning. The meeting will now 3 come to order. This is a meeting of the ACRS Subcommittee on 4 the Advanced Boiling Water Reactor.
5 I am Carl Michelson, Chairman of the ACRS .
6 Subcommittee on the Advanced Boiling Water Reactor. The other 7 ACRS members in attendance are Bill Kerr, Forrest Remick, Paul 8 Shewmon is coming later, David Ward, Charlie Wylie, and our i
9 consultant today is Jesse Ebersole.
10 I will have a few introductory remarks about the 11 Final Design Approval, FDA, and the reason that we're here 12 today in just a moment.
O 13 This meeting will concentrate on the first Review 14 Module consisting of Safety Analysis Report, Chapters 4, 5, 6, 15 and 15 - 1.
16 Richard Major, on my right, is a Cognizant ACRS staff 17 member for today's meeting.
18 The rules for participation in today's meeting have 19 been announced as a part of the notice of this meeting that was 20 published in the Federal Register on May 18, 1988.
21 This meeting is being conducted in accordance with 22 the provisions of the Federal Advisory Committee Act and the 23 Government in Sunshine Act.
24 We have received no written or oral statements from
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25 members of the public. It is requested that each speaker first Heritage Reporting Corporation (202) 628-4888 l
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4 1 identify himself or herself and' speak with sufficient clarity 2 and volume so that he or she can be readily heard.
3 Do any subcommittee members have any initial 4 statements at this time? Seeing none, I would like to proceed 5 with a few introductory remarks to set the platform for this 6 particular series of subcommittee meetings.
7 We are starting today our formal review of the 8 Advanced Boiling Water Reactor and it is for the purposes of 9 issuing a letter for a final design approval and subsequent 10 certification of this particular design.
11 The FSAR that we have, and all subcommittee members 12 have a copy. Is that correct?
13 MR. MAJOR: Those that requested a copy.
14 MR. MICHELSON: Those that requested a copy, okay.
I 15 NOw, the FSAR which most of you have is divided into four 16 modules. I've given a handout whi'.:h is a handwritten one, 17 which I don't seem to have a copy. You have in front of you.
18 It shows the ABWR chapters or volumes that will be in each of 19 these four modules that the staff has divided the work into.
20 On here you will see what date GE is going to submit 21 the particular module and what date the staff will have its SER 22 prepared for that module.
23 At the bottom of the same sheet you will see a 24 listing of eight proposed subcommittee meetings to cover this 25 work over the next two years.
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5 1 You will see the date on which the subcommittee 2 meeting is planned on being held and the subjects of the 3 meetings. -There's a little abbreviation here. The V here 4 means this is when we get our overview of the particular 5 chapter.
6 The plan is to receive of the module and then have a 7 following subcommittee in which we ask more detailed questions 8 and deal with more detailed subjects that the subcommittee may 9 have on that particular module. If they want to see more in 10 depth or whatever.
11 You will see, then, that our work kind of peaks up 12 towards the third and fourth meeting where we will have to 13 write a final letter on one module at the same time we overview 14 another module, and then going to questions perhaps on a third 15 module.
16 To do this we will probably end up with subcommittee 17 meetings that are more than one day of duration during the 18 middle of that activity.
19 Our final meeting is scheduled for about June 1990.
20 It will be when we will prepare the ACRS final letter on the 21 ABWR.
22 Now, the staff is preparing draft final letters on 23 each module as we go along. The subcommittee will also prepare 24 draft letters on each module as we go along, and then there
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25 will be a final integrated letter which we will prepare and a Heritage Reporting Corporation (202) 628-4888
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6 1 final integrated letter which the staff will release.
2 So, that's the schedule. You can see where we're at.
3 It's a pretty fair amount of work, so we need to get about 4 doing it.
5 Now, this is a kind of a new experience for the ACRS 6 because in the past, although we have looked at a number of CP 7 and OL type applications, we have generally not looked at an 8 application so early in the game. We are going in parallel 9 with the stop activity in this case and not following up at the 10 end. This will require more work on our part and more 11 attention on our part, but, hopefully, it will result then in 12 no surprises at the end of the game.
O kl 13 I think it is an interesting &nd challenging 14 opportunity in that we are dealing now with a standard plant 15 design to be certified for use for the next, at least, thirty 16 years. So, we want to enter into it with due care.
17 There are going to be some procedural problems along 18 the way. One of them that's already beginning to appear is 19 that this work by GE is getting out of sync with the work by 20 EPRI on the Approved Light Water Reactors. We were kind of 21 concerned. We don't want to get to the point where we're 22 writing letters on APWR before we're writing letters on 23 comparable chapters of the EPRI design criteria which are 24 supposed to be, in many respects, controlling. Hopefully, they 25 get into sync and they follow along hand in hand.
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t 7 1 We have some concerns about how we're going to 2 handle the severe accidant considerations and this will have to 3 be worked out fairly soon.
4 We also have concerns about how the resolutions of 5 USIs and GIs will be incorporated in. I think this will go 6 along, but view of the problems that we're having with such 7 USIs as A-17 and A-47, we're going to have to get with GE early 8 on to find out how they are approaching their own view of these 9 USIs.
10 Another problem which I have, and I don't know quite 11 how to resolve yet, and that is our subcommittee is almost as 12 big as the full committee. That's no particular problem. The 13 difficulty is, though, what do we bring to the full committee 14 since everybody is at the subcommittee. If there are only one 15 or two people different than full committee, I'm not quite sure 16 how we should we approach this question of how much 17 presentation to make to the full committee.
18 So, you might all want to think about that a little 19 bit and give me whatever recommendations you have because it 20 just isn't clear to me what the best to proceed might be.
21 D P. . KERR: In 1990, the problem may have taken care i
22 of itself.
23 MR. MICHELSON: Well, I'm not thinking of 1990. I am
- 24 thinking more of, you know, the next few months as to what to 25 bring to full committee after having.such a large subcommittee i Heritage Reporting Corporation I
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'L/ 8 1 meeting. So you might want to keep that in mind as we go 2 through. ,
3 Are we scheduled for full committee this time?
4 MR. MAJOR: Just the subcommittee chairman's report.
5 MR. MICHELSON: Okay, just the chairman's report.
6 So, we won't have the problem at least this time. What do we 7 have 30 minutes? 15 minutes?
8 MR. MAJOR: I thought it was 15.
9 MR. MICHELSON: Okay. This time we will not have the 10 problem and maybe we can use part of that time to discuss how 11 we should do it in the future.
12 I believe that just about takes care of what I had in k/ 13 terms of introductory remarks, and you do have the schedule in 14 front of you. So, what we'd like to do today is get started.
15 I beli;ve Mr. Dan Wilkens is going to represent GE and Mr. Dino 16 Scaletti will represent the staff, and I believe the kickoff on 17 the schedule is by GE.
18 (Slides Shown.)
19 STATEMENT OF: MR. DAN WILKENS, GENERAL MANAGER, THE ADVANCED 20 BOILING WATER RCACTOR PROGRAM, GENERAL ELECTRIC 21 MR. WILKENS: Thank you, Mr. Michelson. Let me say 22 it's a pleasure for General Electric to be here today to talk 23 about our Advanced Boiling Water Reactor Program.
24 I am Dan Wilkens, General Manager of the Advanced 25 Boiling Water Reactor Program at General Electric. That Heritage Reporting Corporation (202) 628-4888
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1 includes our activities worldwide in this area.
2 We're here today to try to give you a brief 3 programatic overview of what's going on in the ABWR program, 4 both in the U.S. and Japan.
5 We've brought with us a strong team of GE technical 6 people who will then get into the agenda topics that you'have 7 indicated you want to talk about and will go into the technical 8 areas associated with the first module in some depth.
9 I would like to briefly cover these topics, a little 10 background on the ABWR, and a summary of its key features and 11 why we picked them, talk a little about what's going on in 12 Japan and how that relates to our activities here in the U.S.,
13 give you a summary of our contacts with the Commission and 14 prior ACRS meetings, talk about our certification prog;am here 15 in the U.S. and what we want to accomplish in this program, and 16 then move on to the other agenda topics.
17 We've been here before. So, I don't plan to spend a 18 great deal of background time today, but let me just remind you 19 of a couple of things.
20 The ABWR development program has been going on for i
21 some eight years as an international design effort really to 22 develop an international standard plant. It's a 1350 MW plant
- 23 as it will be applied in Japan. Of course, it could be 24 operated a lower power levels.
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25 It's being done by an international design team, Heritage Reporting Corporation (202) 628-4888
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%] 10 1 initially with participants from Europe, Japan and the United 2 States, and more recently focused more in Japan than the U.S.
3 Our objective with the ABWR was to come up with a 4 plant which would be both advanced and proven. In order to do 5 this, we turned to worldwide BWR technology and tried to look 6 at the BWRs in the U.S. and Japan, in Europe, Italy, Sweden, 7 Germany, and pick the best features that we saw from all the 8 BWRs in the world and put them together in a single design.
9 I'll show you in a moment what the key features are and where 10 they came from, but this is very much an international design 11 put together by an international design team.
12 Our objectives here in the U.S. are two. We are, as 13 you know, working very closely with EPRI on the ALWR 14 requirements program, and our objective is to adapt this design 15 as necessary to meet the EPRI requirements.
16 As we see the future, the EPRI requirements program 17 in effect is writing what we hope will be the utility bid 18 specification for future plants so that any utility wanting to 19 order a plant in the U.S. in the future would come out with the 20 EPRI requirements document rs a Bid Spec.
21 We, obviously, would like to be in a position with 22 the ABWR of being able to provide a design that meets that 23 specification. So, we're working very closely with EPRI and,
(~T 24 in fact, have made some adjustments to the design already in
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25 order to come into line with the EPRI requirements.
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L/ 11 1 We, also, are very eager to see the-design that is 2 certified here in the U.S. be the same design that we are 3 building in Japan so that we will be, at the end of this 4 program, in a situation where we can say here is a certified 5 design and, in fact, the lead plants being built in Japan are 6 that design and so the certification is, in effect, backed by a 7 live project that in the mid-90s will be producing power.
8 Our objectives technically have been to improve 9 essentially all areas of plant performance and economics.
10 Improved safety and reliability, operability, capacity factor, 11 and reduction in all elements of plant cost. We think the ABWR 12 design had, in fact, achieved all of those things and achieved 13 them all at once. ,
14 I mentioned that it's an international design by an 15 international team and let me just highlight some of the key 16 features in the ABWR and where they'came from.
17 The core and fuel design is following the General 18 Electric fuel product evolution. We did not, and do not intend 19 to, have a unique core and fuel design for the ABWR, but rather i 20 as we advance the BWR fuel design through successive production !
i 21 evolutions, we are designing the ABWR so that it will be able 22 to accept those advancing concepts the same as all of our 23 operating BWR plants will be able to accept them.
{) 24 25 In the re-circulation system, we studied the various designs around the world and eventually settled on the internal Heritage Reporting Corporation (202) 628-4888 4s , , , . - - - - . - , . . . - - . . , , , - -
/'s, U 12 1 pump re-circulation design along the lines of what has been 2 done in Swedish plants.
3 In our case, ten internal pumps mounted at the bottom 4 of the vessel. They are wet motor pumps so that they do not 5 have the seal problem to contend with. We find that this 6 feature has many attractive features. One of which is to 7 greatly reduce the design base's loss of cooling action by 8 eliminating by large piping and, in fact, by eliminating the 9 large Recrc. LOCA which has always been the key design basis 10 LOCA in the older BWRs.
11 MR. MICHELSON: Question for the gentleman? Do the 12 BWRs in Sweden all have internal Recrc. pumps?
13 MR. WILKENS: That is correct. The second feature 14 I'd like to highlight is the fine motion control rod drive.
15 This is a drive that scrams hydraulically, is maneuverable 16 electrically and, in fact, can provide an electrical backup t 17 the hydraulic scram providing an added degree of diversity.
18 This was a design that was originally developed by 19 General Electric many years ago, but at the time we never 20 applied it. It has been applied in Germany and is used on all 21 of the current German BWRs. We're, in effect, bringing it home 22 by applying it to the ABWR.
23 It has the attractive features of diversity. It
{) 24 25 eliminates the scram discharge volume which has always been one of the controversial areas in the BWR design. We think it is a l
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(_) 13 1 real improvement.
2 We've gone to an advanced pressure suppressing 3 containment which is a modification of our MARC III design.
4 It's a reinforced concrete containment. It has horizontal 5 vents like the MARC III, but it has the wet well, the' 6 suppression pool, in effect, covered in the wet well and the 7 containment can, therefore, be inerted like the older MARC Is 8 and IIs which provides, we think, a better approach to the 9 hydrogen issue.
10 In addition, we have structurally integrated the 11 containment in the reactor building so that rather than having 12 the free-standing containments we've had in many plants, we've 13 structurally integrated the containment in the building which 14 provides a greatly increased seismic capability.
15 The service and maintainability features in this 16 plant have been adopted from the Japanese approach. The 17 Japanese have really done a magnificent job in eng!.neering 18 their plants with service and maintenance factoren into the 19 design from the very beginning, and we've taken a chapter out 20 of their book and applied that whole approach.
21 All of the major equipment in this plant has s plan 22 for service and maintenance that includes the necessary 23 handling equipment, the necessary space, and in many cases 24 service facilities located nearby for performing the required
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{JT 14 1 Finally, the design employs solid-state digital 2 control systems for both safety and non-safety systems. Along 3 with the multiplexing, to eliminate a lot of the wire and cable 4 that we've historically had in our plants.
5 These control systems are self-diagnostic. They are 6 continually monitoring themselves. They tell you when failures 7 occur, where they've occurred, and there's enough redundancy in 8 the design so that you can, in fact, go down and remove.the 9 failed card and replace ic with a new card without interrupting 10 plant operation or causing a scram in the process.
11 We think that this is a great improvement in terms of 12 the reliability of the plant.
13 MR. MICHELSON: That particular new technology 14 feature, of course, means that you'll have to pay close 15 attention to the environmental cooling of that equipment.
16 MR. WILKENS: That's right.
17 MR. MICHELSON: And you're going to tell us, as time 18 goes on, what provisions you've made'for that?
19 MR. WILKENS: Absolutely.
20 MR. MICHELSON: Thank you.
21 MR. EBERSOLE: I'd like to ask a question. With that 22 slide up there, and I hope I can do this clearly. We're in the 23 shadow of the Shoreham tragedy, as far as I'm concerned. I 24 would like to think this plant could be placed where Shoreham 25 is and there would be no problems, and a new look would be Heritage Reporting Corporation (202) 628-4888
15 1 taken at emergency procedures of vacating the scene versus 2 sheltering, and a new set of concepts would be available for 3 the openness of your accident consequence if you can furnish 4 that.
5 I take it this design will look, although it doesn't 6 have to do so in foreign countries, I guess, Japan, et cetera, 7 as to what to do about it to minimize this evacuation concept, 8 but you will have some figures that you'll tell us about how 9 you would at least attempt, ideally I would say you could dig 10 up Shoreham and put this plant there and we wouldn't have a 11 problem.
12 MR. WILKENS: Well, I won't comment on who might be 13 brave enough to put another plant at Shoreham right now, but 14 the plant does have features which we believe represent a step 15 forward in both prevention and mitigation of severe accidents.
16 When we worked with the NRC staff over the past year 17 to develop the licensing review basis for this plant we have, 18 in fact, committed to some goals in terms of probability of 19 core damage and probability of events that would lead to off-20 site consequences which are quite ambitious, and I'll, in fact, 21 mention those in a minute.
22 Now, I mentioned that the ABWR is Ipoving forward in 23 Japan. Last year the Tokyo Electric Power Company announced 24 plans to proceed with the two lead ABWR units in Japan on the i 25 schedule that I've shown here.
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O 16 1 In fact, the licensing application was formally 2 submitted to the Ministry of Trade and Industry last Monday for 3 those two lead units. We expect the licensing to be complete 4 in about two years. The project will then move on into 5 construction, with the first plant scheduled for commercial 6 oper,ation in '96. The second one, in '98.
7 We were particularly pleased that later last year 8 Tokyo Electric Power Company announced that those plants would 9 be built by a joint venture of General Electric, Hitachi, and 10 Toshiba, and within that .cint venture they selected General 11 Electric to provide the nuclear steam supply systems, the fuel,,
12 and the turbine generators for those lead units.
O 13 So, this is not just a paper plant or at least it's 14 not going to be a paper plant for very long. This is, in fact, 15 a real project that is moving forward in Japan on a very 16 aggressive schedule and will be the next generation boiling 17 water reactor in Japan.
18 Of course, as part of the very successful Japanese 19 nuclear program which, as you know, is very much founded on the 20 concepts of standardization that we're trying to implement 21 here.
22 MR. MICHELSON: On the schedule you've shown there, 23 is that allowing about 48 or 52 or how many months for
(} 24 construction? -
25 MR. WILKENS: The construction schedule from first Heritage Reporting Corporation (202) 628-4888 I
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1 concrete to commercial operation will be 48 months.
2 MR. MICHELSON: For commercial operation?
3 MR. WILKENS: Yes. First concrete to commercial 4 operation. In fact, TEPCO is putting some pressure on us to 5 shorten that. The Kashiwazaki site is a rocky sea coast out on 6 the Japan sea side of Japan. There's a fair amount of time in 7 that schedule for excavation of that site which takes on the 8 order of 18 months at the front end before any real plant work 9 is done.
10 MR. MICHELSON: Is that in the 48 month schedule, 11 too?
12 MR. WILKENS: No.
13 MR. MICHELSON: Or that's ahead of time?
14 MR. WILKENS: No, that's ahead of it. I said from 15 first concrete to commercial.
16 MR. MICHELSON: Yes, I see.
17 MR. REMICK: Can that early work be done before 18 licensing?
19 MR. WILKENS: No.
20 MR. REMICK: No?
21 MR. WILKENS: No. That will be done after the 22 establishment permit is issued which is in about this time in 23 1990. Two years from now.
24 MR. REMICK: Looking to what GE will supply for those 25 two units, what differences if any are there in the Heritage Reporting Corporation (202) 628-4888
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s- 18 1 requirements for the quality of components? Will you supply 2 what you would for a U.S. plant?
3 MR. WILKENS: I would say the quality assurance 4 requirements in Japan are very similar to what they are in the 5 U.S., and if we built-a plant in the U.S. our objective would 1 6 be to build the same plant that we would be building in Japan.
7 MR. REMICK: The same components?
8 MR. WILKENS: We would like to keep sources of supply 9 flexible but, basically, yes. Certainly all the compunents 10 would be bought to the same specifications that we would use in 11 Japan, and we would want to buy components from qualified 12 suppliers. Obviously, when we're finished with these two 13 plants we will have at least some qualified sources of supply.
14 MR. REMICK: Would you have to supply the same QA 15 pedigree for those components in Japan as you would for a U.S.
16 plant?
17 MR. WILKENS: Yes.
18 MR. EBERSOLE: What detail do they require of you in 19 definition of components and pieces of equipment? Right down 20 to catalogue numbers?
21 MR. WILKENS: Eventually, Japanese regulatory process 22 will look down into the details of component design and stress 23 analyses and so forth. I would say that establishment permit process, which is going on in the next two years in Japan, is
{} 24 25 comparable to the certification process that we are looking at i
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2 It's basically system design, lay out and 3 arrangement, and down through the specification of the 4 equipment, down through, in effect, the procurement level of 5 definition.
6 MR. WYLIE: Will the components be built in Japani 7 MR. WILKINS: Some will. Some will not. We plan to 8 build some of the turbine generators, for example, primarily in 9 S'chenectady, here in the U.S. We will build some of the ,
10 reactor components in our Wilmington plant. We intend to 11 source a lot of equipment in Japan.
12 MR. EBERSOLE: Where will you build the control rod O 13 drives?
14 MR. WYLIE: They will be built in Japan.
15 DR. KERR In response to an earlier statement you 16 said that you expected the components and equipment would be 17 built to the same set of specifications as a U.S. plant's.
18 That's a meaningful statement, but it doesn't necessarily 19 indicate the quality of the components.
20 Do you expect any significant difference in quality 21 of components built in Japan and in the U.S.?
22 MR. WYLIE: Well, I think you really have to address l 23 that question on a component by component basis. I would say,
{} 24 today, when you look at the ABWR, for example, the fine motion 25 drive. Right new that drive is not built by anyone in the U.S.
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20 1 It is built by, for application in Germany and you can 2 certainly get them from Craftwork Union.
3 We have worked with Hitachi and Toshiba and are in 4 the process of qualifying both of those companies as suppliers 5 of that drive. In fact, we have one of those drives that we 6 have done a lot of testing on at San Jose. We, right now, have 7 one operating in the LaSalle plant with Commonwealth Edison in 8 plant test.
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9 So, we would say that are at least, or will be by the 10 time this plant is built, at least three qualified sources of 11 supply for that drive. Whether we in the future develop more 12 or decide to make it ourselves is a future decision, but for O 13 all of these components our approach would be to buy them to 14 these approved specifications and only from qualified suppliers 15 who have shown they can build them and then put them through 16 testing.
17 DR. KERRt But in this country with a very elaborate 18 QA system and what I would presume are reasonable bid 19 specifications, or whatever, we find equipment in nuclear power 20 plants supplied by U.S. suppliers that are substandard in many 21 ways. '
22 My impression is that is less likely to be the case 23 for components and equipment supplied by the Japanese.
(} 24 MR. WYLIE: I really think the difference is one of 25 standardization and qualification as opposed to U.S. versus Heritage Reporting Corporation (202) 628-4888
21 1 Japan.
2 The Japanese have done a_ magnificent job of 3 standardizing, qualifying equipment and then standardizing it.
4 That's exactly what we're trying to do here with this design 5 and the certification program, is get in the same' situation.
t DR. KERR I just hope that we don't sta.dardize on 7 the basis of some of the equipment that has gone into U.S.
8 plants.
9 MR. WYLIE: I wouldn't argue with you.
10 MR. MICHELSON: Why don't we proceed.
11 (Continued on next page.)
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n U 22 1 MR. WILKENS: This chart compares the licensing 2 schedules in the U.S. and Japan and the main point is they are 3 on virtually identical schedules.
4 We began a little earlier here in the U.S. with the 5 formal licensing process, but as you know we are going through 6 this modular process to keep in step with EPRI and in Japan the 7 establishment permit was submitted last Monday and is scheduled 8 for issuance two years later.
9 MR. MICHELSON: That is not the same slide that you 10 gave us in the handout. Would you provide us a copy of that 11 one?
12 MR. WILKENS: Yes. Is that different?
13 MR. MICHELSON: Yes.
14 MR. WILKENS: Okay. We will you provide you with 15 copies.
16 MR. MICHELSON: It is a good slide, too. They are 17 both good. This is more complete because it goes on further in 18 time.
19 MR. WILKENS: Yes. We will give you a copy of this.
20 MR. MICHELSON: Because this is important 21 information.
22 MR. WILKENS: Let me emphasize a point while I have 23 this up. We very much want to keep what we have built in Japan
[} 24 25 and what we certify here in the U.S. the same and our Japanese associates very much want to keep it the same. And so it is Heritage Reporting Corporation (202) 628-4888
23 1 . going to be a continuing challenge through this program.
2 .The requirements in a few areas in Japan are a little 3 different than they are in the U.S. and case-by-case we are 4 deciding whether to meet both or whether to meet one and argue 5 it ought to he applied in both places, and those kind of issues 6 are going to come up from to time.
7 Our Japanese associates also very much want to hold 8 this schedule, because they are moving on into constructing a 9 plant. They would not like to be surprised after they begin 10 building to find that we do something different here in the 11 U.S. and end up not meeting our objective of having an 12 international standard.
13 And so we are very dedicated to keeping on this 14 schedule and as you carry out your review, you will find that 15 GE will be doing everything possible to try to provide you the 16 information that you need so that we can keep up with this 17 schedule.
18 MR. MICHELSON: It is appreciated, of course, that 19 this is a U.S. review.
20 MR. WILKENS: Oh, yes.
21 MR. MICHELSON: And will have to meet U.S. standards 22 whether the U.S. deal with it the same or not.
23 MR. WILKENS: We may from time-to-time try to 24 convince you that a Japanese standard which we are applying in
}
25 Japan may be suitable here in the U.S. in lieu of what we have Heritage Reporting Corporation (202) 628-4888
( .
24 1 historically done in the U.S.
-2 Our feeling is, when we think standardization we 3 should really have our sights set on the world, not on the U.S.
4 And the nuclear enterprise is a world enterprise. There are
, 5 good ideas from overseas, and we think that it is well worth to 6 tzar to achieve a design that can be built anywhere in the 7 world.
8 DR. KERR: Well, I think we all agree that 9 standardization has some virtues, but one of the vices it can 10 have is that one might standardize on the wrong thing.
11 MR. WILKENS: Right.
12 DR. KERRt And so we all want to be careful that the 13 standardization is a standardization on what should be 14 standardized.
15 MR. WILKENS: No question. But let me say that the 16 approach that is being followed here I think really addresses 17 that in the sense that we have been building BWRs around the 18 world for some 30 years now and what we have done with this 19 program is go to those plants all around the world and try to 20 put the best features of all of them together into a next 21 generation and standardize on in effect a plant that has 22 already been built with proven features.
23 MR. EBERSOLE: Can we pick a point that illustrates 24 that? I will pick the point of dependency on AC power.
25 A couple of years ago we were told that the concept Heritage Reporting Corporation (202) 628-4888
25 1 of low imaging voltage power failure in Japan was virtually 2 zero. They never had any failures of diesels to start and they 3 had special machines.
4 I believe this plant has attempted to reduce 5 dependencies on continuity of AC power to some extent. I hope 6 a great deal.
7 Could you comment on for instance? Do the Japanese 8 plants use special machines, low voltage equipment and diesel 9 generators?
10 Or have they gone the way of reducing dependencies on 11 AC power? That is one of our problems, as you know on the wall 12 there, is dependency on AC power.
13 Could you comment on what has been done in the 14 general context on AC power dependency?
15 MR. WILKENS: Well, I think maybe it would be better 16 to have Craig Sawyer talk about that one when he gets up.
17 MR. MICHELSON: Well, later on we can cover that. It 18 would be better to cover it when we get to that particular 19 point. It is the kind of a point that I think we should 20 emphasize.
21 MR. EBERSOLE: Oh, yes. There is quite a few things 22 going by right now that will get emphasized later.
23 MR. REMICK: Dan, you indicated that along the way 24 you might try to convince us that a Japanese design or method 25 is suitable for application in the U.S.
Heritage Reporting Corporation (202) 628-4888
26 1 But I would urge you, I hope if you have something 2 that you think is better than Japan, not only suitable but 3 better, to try to convince us that it is better. Because I i 4 think you are in the unique position of working closely with ,
5 the Japanese. We know they have some good practices that might 6 better than ours.
7 MR. WILKENS: Yes.
8 MR. REMICK: And we should know about them.
9 MR. WILKENS: Yes, and that is exactly what we intend 10 to 5: We have convinced them to adopt many U.S. practices and 11 occasionally they conrince us to adopt Japanese practices, and 12 in cases where they have done that we will bring them here.
O' 13 Let me just quickly summarize where we are in the l
14 certification program. The NRC staff issued a licensing review 15 basis for the ABWR last year. It established some of'the key 16 acceptance criteria. Talked about many of the procedural i 17 aspects of licensing. Laid out the schedule, the modular 18 submittal approach.
19 We reached agreement on the level of design detail 20 that would be provided. Worked out in concept of how the ACRS 21 review would fit into this modular approach and defined at 22 least in basic terms the certification process that would .
t 23 follow the FDA.
24 A key area in there was the area of severe accident 25 criteria and we have committed during this review to show a ,
Heritage Reporting Corporation (202) 628-4888
27 1 core damage probability of less than 10 to the minus 50 per 2 year and a frequency of exceeding 25 REM off site of a decade 3 lower or less than 10 to the minus 6 per year.
4 We also in that document laid out in some detail what 5 it is the staff needs to see in this advanced electronic area 6 in order to conduct their review in that area, and we will be 7 providing that information.
8 MR. REMICK: Dan, could you tell us what your 9 definition of core damage is associated with 10 to the minus 5?
10 It is something that we all throw around loosely. You are not 11 saying core melt there, I notice. You are saying core damage.
12 But what is the definition _n your case?
() 13 MR. WILKENS: Craig, do you want to answer that?
14 ,
MR. SAWYER: I am Craig Sawyer from GE. The way we 15 are running the PRA, technically we will consider it to be a 16 core damage event if the peak temperature of the fuel is in the 17 neighborhood of 2400 degrees.
18 MR. REMICK: Thank you.
19 MR. WILKENS: The safety analysis report --
20 DR. KERR: Excuse me. I would like to pursue this 21 general area a little further.
! 22 You said you had committed to demonstrating that the 23 core damage frequency is less than 10 to the minus 5 per year.
"g 24 Does that mean that according to the NRC's PRA or your PRA or
{%)
25 some compromise -- what is the significance of that statement?
Heritage Reporting Corporation l (202) 628-4888 l -- _ _ _ _ _ _ _ . _ . _ _ , . _ _ _ _
.f)
- ~j 28 1 MR. WILKENS: That is in effect an acceptance 2 criteria. We have to convince the staff that it is Inss than 3 10 to the minus 5 per year.
4 DR. KERR: You also said earlier, I think, that you 5 believed that these plants were improved and were safe and more 6 reliable. Does that statement imply that core damage frequency 7 you expect to achieve here is less than the core damage i
8 frequency that is now being achieved in the U.S.?
9 MR. SAWYER: Yes.
10 MR. WILKENS: I think in general, yes, yes.
11 DR. KERR: What do you think is an appropriate number 12 for existing reactors?
h 13 MR. WILKENS: Well, I guess I wouldn't want to 14 comment on that. I think that the approach we have picked here 15 is --
16 DR. KERR: I am just trying to understand how much 17 improvement you expect to achieve with this system. Now I 18 realize that in order to do that you have to make some judgment 19 as to what existing plants are doing, but unless you have made 20 that judgment, I don't see how you can determine whether you 21 have made an improvement or not.
22 MR. WILKENS: Well, the 10 to the minus fifth number 23 is the number that comes out of the EPRI requirements document.
{} 24 DR. KERR: I am going back to your earlier statement 25 which was that you are trying to achieve improvements. Maybe l
Heritage Reporting Corporation (202) 628-4888
/~~'s V 29 1 you didn't mean improvements in this number. If you didn t, I 2 will accept that.
3 MR. WILKENS: No. I think -- look, the number that 4 was in the EPRI requirements document was picked after having 5 looked at the PRAs that have been done on existing plants in 6 this country, and as you know there is qui,te a range of core 7 damage probabilities and many of them are done to different 8 methodologies and it is awful hard to pick a number and say 9 this is the number of existing plants.
10 DR. KERR: If you want to avoid this question, I 11 understaad it, and if that is what you propose to do I won't 12 pursue it any further, r~.
13 I thought perhaps you had indeed said this is the way 14 things are and this is the way we want them to be and this 15 represents improvement. If you haven't done that, I will 16 accept that answer and go on.
17 MR. SAWYER: Bill, we have taken a crack at that, and 18 as well will hear from -- unfortunately, it will be towards the 19 end of the day when we start talking about Chapter 15.
1
! 20 DR. KERR: I promise to stay awake.
l 21 MR. SAWYER: But if you just step back a moment and l 22 forget the numerics of the PRA and look at some feature, the 1
l 23 design changes that we have done to the scram system to improve 24 our reliability for the anticipated panzione without scram.
l 25 The third division for heat removal. A third high Heritage Reporting Corporation (202) 628-4888
. _ _ _ . . _ _ - . _ _ a
I~
( 30 1 pressure injection system. Elimination of any large piping 2 connects to the vessel below the core elevation.
3 Even before you start turning the crank, you have to 4 believe that you have substantially reduced the risk of core 5 damage from those things.
6 DR. KERRr If we are operating on faith -- I am a 7 great believer in faith -- but if we are operating on PRA, then 8 we are operating on PRA which does involve a good bit of faith.
9 I am just trying to understand how you decide on 10 improvements? If it is engineering judgment based on the fact 11 that you improved it?
12 MR. WILKENS: No, we did apply PRA judgments in 0- 13 coming up with the ECCS configuration, for example, that we 14 have in this plant. That was part of the rationalization of 15 the ECCS design for this plant.
16 The reason why you are receiving somewhat of a hedge 17 from us is that there aren't two PRAs out there that have been 18 done exactly to the same set of ground rules. And we have 19 taken a crack at comparing our plant against the existing 20 standard plant in Japan, the BWR-5 standard in Japan, which is 21 similar in some respects, in many respects, to the BWR-5s in 22 the U.S.
23 And that the basis for our rationalizing the ECCS
{) 24 25 part of the system, for example. But it is difficult to make a quantitative judgment of, on a factor of X better than the Heritage Reporting Corporation (202) 628-4888
e
. by 31 1 previous plants. And so what we have done is used a 2 combination of probability analysis kind of thinking and 3 deterministic kind of thinking. What new features have I put 4 in here which are going to make this problem go away, for 5 example, as the basis for our judgment that we have made this 6 plant safer?
7 MR. SAWYER: In your estimate, does your goal include 8 external events?
9 MR. WILKENS: It certainly events. It is external 10 and internal.
11 MR. SANYER: A lot of the PRAs, of course, have not 12 treated external events very effectively. I assume you will do n
(-) 13 it effectively and still maintain these goals?
14 MR. WILKENS: We are giving it our best shot.
15 MR. SAWYER: Yes.
16 MR. WILKENS: One of the areas that we are addressing 17 right now amongst ourselves and also with the staff is how to 18 apply a thinking of external events, particularly earthquakes, 19 with the potential range of site that you apply this plant at.
, 20 That kind of thinking is going to be cranked into risk l
l 21 assessment.
22 MR. SAWYER: But you will include all external l 23 events?
24 MR. WILKENS: All external events. Well, all 25 significant external events.
l l Heritage Reporting Corporation j (202) 628-4888 i
32 v
1 MR. SAWYER: Well, you don't know if they are 2 significant until you look at them carefully.
3 MR. WILKENS: Well, we are talking about fires, 4 floods, earthquakes, tornadoes, that kind of thing.
5 MR. SAWYER: Yes. Okay, not airplane crashes.
6 MR. WILKENS: We are not going to worry about 7 meteorites and some things that are really low probability 8 events.
9 MR. EBERSOLE: Let me ask a question. One of the 10 contributors to your problem is the ATWS event? It seems to me 11 about three or four months ago, I saw an analysis -- I am not 12 sure where it was, who did the work, Brookhaven, I think -- but
() 13 my impression was that you had made a finding that the 14 thermohydraulic neutronics consequences of an ATWS is a great 15 deal less than it has been heretofore thought.
16 Am I correct about that, Bill? You might know about 17 that. The ATWS consequence is not as vigorous as it used to be 18 thought?
19 DR. KERR: Well, I have seen some that say it is not 20 as bad, and some that say it is worse, Jesse.
21 MR. EBERSOLE: I don't know where we are in that 22 context.
23 DR. KERR: Take your pick.
24 MR. WILKENS: I would like to suggest may be hold on 25 that until Frank talks a little later?
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l l Heritage Reporting Corporation i
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i 33 1 MR. MICHELSON: Yes. We will get to ATWS quite a few 2 times before we are done.
3 MR. WILKENS: Well, this summarizes where we are in 4 the module submittal process. There are two modules already 5 in. The third one is due in at the end of June and they will .
6 all be in by the end of this year.
7 So that is a very quick overview of our program.
8 MR. MICHELSON: Let me ask you -- you said they are i
9 in by the end of the year. Is that including the PRA?-
10 MR. WILKENS: Yes.
11 MR. MICHELSON: So you are a little ahead of 12 schedule, or you are certainly on schedule.
() 13 MR. WILKENS: We are on schedule. I would be 14 hesitate to say we are ahead.
15 DR. KERR: I have a question. This may not be the P
16 time for an answer, but tell me where it should be asked.
17 Do you have a quantitative reliability or other goal 18 for your non-safety control systems?
19 You were talking about multiplexing earlier as a 20 desirable feature and it is certainly a desirable feature, but 21 I was curious as to whether you had set any goal for 22 performance, internal goals.
23 MR. SAWYER: Bill, there are goals. Forgive me for
,_ 24 not knowing the exact number, but we have written quantitative V 25 goals for the multiplexing system.
Heritage Reporting Corporation (202) 628-4888
34
\~s .
1 MR. MICHELSON: For non-safety multiplexing?
2 MR. SAWYER: Yes, the control systems and the safety 3 systems both.
4 MR. MICHELSON: We will get to those eventually.
5 MR. WILKENS: Let me just finish by emphasizing the 6 fact that we think this program is really a unique opportunity 7 to put the U.S. back on its feet in terms of doing what we 8 probably should have been doing many years ago, namely, 9 certifying standard plants in advance and then building them 10 all in the same way.
11 And with the ABWR moving forward in Japan and the 12 certification program here in the States, we think we have
() 13 really a very unique opportunity in the next two years to 14 achieve a standardization program here in the U.S. And we are 15 very dedicated to seeing that happen.
16 MR. MICHELSON: Could I see your previous slide a 17 moment? I am having trouble with the dates on it, and maybe 18 you corrected them on your slide.
19 MR. WILKENS: Well, I didn't. Here, let me give you 20 the code.
21 MR. MICHELSON: Okay.
22 MR. WILKENS: The first one is '87. All of the rest 23 of them are '88.
7, 24 MR. MICHELSON: Those are all like 3-30-88.
25 MR. WILKENS: All the rest are like 3-30-88.
Heritage Reporting Corporation (202) 628-4888
() 35 1 MR. MICHELSON: Okay.
2 MR. WILKENS: Thank you. Let me introduce just 3 quickly the members of the GE team that are here. You have 4 heard from Dr. Sawyer, already. He is our Manager of System 5 Design and Performance Engineering on the ABWR.
6 Joe Cork, who is sitting to his left, is the Program 7 Manager'for our U.S. ABWR programs including the certification 8 program.
9 Gentry Wade, who will talk next, is our Manager of 10 Containment and Configuration Engineering.
11 Irv Kobsa, who is sitting to his right, is our 12 Principal Engineer on Reactor Design. He will be talking to
() 13 you about the reactor resert control rods and so forth.
14 Gentry?
15 (Continued on next page) 16 17 18 19 20 21 22 23 24 25 Heritage Reporting Corporation (202) 628-4888
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(/
1 MR. MICHELSON: Before you sit down: Who handles 2 the real project coordination from the licensing and general 3 coordination with NRC? Are you the gentleman that will be 4 doing that?
5 MR. WADE: Well, Mr. Cork is our program manager and 6 we work through our licensing organization.
7 MR. MICHELSON: Bill is the fellow who really is 8 doing the detail work on coordinating licensing.
9 MR. WADE: Yes.
10 You asked for a brief presentation of the ABWR
^
11 containment, specifically, the arrangements that are ins 1de the 12 containment. We will go over that. Primarily, I think we can
() 13 make major use of this slide, this isometric.
14 (Slides shown.)
15 Some of the objectives we set out for the arrangement 16 of the ABWR reactor building which you see here in total and 17 the containment which is enclosed in the dark gray line, the 18 objectives were to provide a clean and controlled access to 19 this reactor building. If you go into a clean area to diesel 20 generators, electrical rooms, you are always in a clean area.
21 There is no contamination involved.
22 If you go in any of the controlled areas, you can 23 eventually get inside the containment, you can get to the 24 refuelling area, you can get to the ECCS equipment rooms and b)
25 that sort of thing.
i I
Heritage Reporting Corporation (202) 628-4888
7 37 1 As Dr. Wilkins mentioned, the Japanese are very 2 strong on servicing and maintenance. You will find that we 3 work very hard to get a building that you could walk completely
, 4 around on your daily patrols, 360 degree access in the reactor 5 building, 360 degree access inside the drywell. So, it 6 improves the maintainability of the plant which eventually has 7 an impact on safety.
8 We put in servicing monorails to handle equipment.
9 There are special equipment rooms right outside the hatches to 10 the drywell where you can work on the fine motion control rod 11 drives and the pumps, that sort of thing.
12 This is a pressure suppression containment as all BWR
(]) 13 recently have been. It is a cylindrical reinforced concrete 14 containment. It has typical steel drywell heac'.. It is lined 15 with steel plate, stainless steel in the wetted tection of the 16 suppression pool. And it is integrated with the surrounding 17 reactor building, the floors of the reactor building. And, of 18 course, the fuel burners on top that go across the drywell 19 head. It is a completely integrated structure.
20 We will talk about this in mare detail later on. It 21 has a horizontal vent system identical in configuration to the 22 Mark III as far as spacing and size of vents goes. We will 23 talk more about that. But you have a vent path from the upper 24 drywell, down through the pedestal into the suppression D
d 25 chamber.
Heritage Reporting Corporation (202) 628,-4888
38 1 The design pressure on this containment is.45 psig.
2 MR. EBERSOLE: May I ask a question?
3 MR. WADE: The old Mark III containment avoided one 4 of the containment bypasses problem by requiring the lines from 5 the SRVs to remain in the drywell until they got into concrete 6 and went under the water.
7 I think this design, the pipes from the SRVs covered 8 the wet well aspects. Am I correct?
9 MR. WADE: They do.
10 MR. EBERSOLE: So, what is the rationale for 11 guaranteeing you can't have being with bypass a rupture of 12 these pipes?
13 MR. WADE: The SRV discharge piping is built to a
(])
14 high quality, higher quality than normal.
15 MR. EBERSOLE: The Germans used double wall pipes.
16 Would you intend to do that?
17 MR. WADE: No.
18 MR. EDERSOLE: So, you now have returned to this old 19 argument about SRV pipe failure and the quality of the pipe 20 being good enough to argue that that will never occur.
21 MR. WADE: Yes.
22 MR. EBERSOLE: And you have to brace it, et cetera, 23 et cetera, et cetera.
24 MR. WADE: It has supports, yes.
l
( 25 MR. MICHELSON: The last time we heard a f
Heritage Reporting Corporation (202) 628-4888
r'S 39 O
1 presentation, I think you were still using concentric piping on 2 the SRVs. Now, you are saying you have gone back to single 3 wall pipe. Is that right?
4 MR. SAWYER 4 If I remember the conversation, we said 5 we were going to convince the staff that the stresses on the 6 discharge piping would be acceptably low.
7 MR. MICHELSON: Maybe I don't remember it the same, 8 then.
9 MR. EBERSOLE: Was it impossible to continue to use 10 the practice in a Mark III containment which was to eliminate 11 that question?
12 MR. HADE: Yes, it is in this configuration.
(} 13 MR. EBERSOLE: It is. You just simply can't do that 14 anymore?
15 MR. WADE: No.
16 MR. EBERSOLE: You mean you can't use concentric 17 piping anymore?
18 MR. WADE: You can't get down between the reactor 19 vessel and the skirt and the pedestal and come out in the 20 suppression chamber in this. I think that will be more evident 21 in a later slide.y 22 MR. EBERSOLE: Okay.
23 MR. WADE: The seismic capability of this plant for 24 U.S. standards are .3 g for the SSE. It's much higher in
) 25 Japan, however, their seismic inputs are different. One Heritage Reporting Corporation (202) 628-4888
40
- (~]
1 important thing to remember in the design is that we have a 2 four-quadrant separation for the electrical C&I systems. Each 3 quadrant of the building, starting with the reactor steamlines 4 is a 'ifferent separation quadrant. The mechanical, because we 5 have three CCS systems and three emergency diesel generators, 6 there are three different quadrants in the building.
7 MR. EBERSOLE: If you can qualify what you mean by 8 separation in the context of the strength of the concrete or 9 whether it is concrete and what are the penetrations and 10 whether they are HMV duct works. When you say, "asparation,"
11 it really, really conveys very little meaning as to what you 12 mean by a separation in the sense of degree.
(} 13 14 MR. MICHELSON: Well, Jesse, we are going to have a whole discussion later on physical separation. Layout is going 15 to be a very important consideration.
16 MR. EBERSOLE: Okay, so that will be a topic in this 17 other room.
18 MR. MICHELSON: And then we are going to hit the 19 separation very hard from the environmental viewpoint and 20 external vent viewpoint and so forth.
21 We will give them a forewarning of what we would like 22 to hear about.
23 MR. WADE: We are going to talk now about the 24 different areas in the containment inside the containment.
() 25 There is the upper drywell, which is this cylindrical area Heritage Reporting Corporation (202) 628-4888
41 0 1 here. There is the lower drywell which is the area under the 2 reactor vessel. These are connected together by a series of 3 drywell connecting vents which will be more evident in a 4 picture we show later on.
5 The suppression chamber is a cylindrical chamber on 6 the outside. This is really sort of an over-under containment 7 as you might be used to in Mark II in that the suppression 8 chamber is directly under the upper drywell. It has cool water 9 in the bottom and the air in the top.
10 The difference'in this design, one of the primary 11 differences in this design from previous designs is we have 12 tried to improve access to the drywell.
rg 13 MR. REMICK: Question?
U 14 MR. WADE: Yes.
15 MR. REMICK: Is there any venting capability of the 16 airspace above the suppression pool?
17 MR. WADE: No. There is no venting airspace. There 18 is a HVAC connection to that area that you use for inerting and 19 de-inerting, but there is no specific venting path.
20 MR. EBERSOLE: You cannot revert to the beautifully 21 simple process that you propose on VSAR called the UPPS 22 approach which I think the public might believe in?
23 DR. KERR: Excuse me. I didn't understand what 24 approach?
() 25 MR. EBERSOLE: Well, the ultimate advantage even Heritage Reporting Corporation (202) 628-4888 8
7
42 g
V) 1 which the man on the street can understand of a boiler is that 2 when you get in duress, you can boil it open to atmosphere if 3 you don't have damaged fuel. And that was a main feature to 4 cover. virtually all potential kinds of accidents in your last 5 design before this one. And is it provision of blowdown, the 6 initial condensation of steam in the suppression pool and 7 eventually evaporation of the suppression pool to atmosphere?
8 In short, the core would revert to no more than a potful of 9 pins. And in the brutally simple concepts boiling to 10 atmosphere with any source of low pressure water that you can 11 find to put on the core. That's the kind of a thing thi; 12 people might even believe in if you had a good shoreham. But I 13 really am disturbed to see that that has gone by the board as 14 well as this bypass problem on these SRVs.
15 MR. MICHELSON: Well, when we get to the severe 16 accident discussions, we are going to discuss at great lengths 17 what their plan is to handle the severe accident case and then 18 they'will have to explain why they don't need a last ditch kind 19 of --
20 MR. EBERSOLE: Public acceptance is a major aspect of 21 these plants. And that doesn't mean you can do it with complex 22 systems. .
23 DR. KERR: Le make sure I understand what the
, 24 situation is. There is not specific provision for venting in j
() 25 this containment.
Heritage Reporting Corporation i (202) 628-4888
43 O 1 MR. WADE: That is true. There is no specific system 2 that I would call a venting system that they are presently 3 putting in some of the reactors in Europe, which is that.
4 There is a ventilation duct that goes out. The radiation 5 levels in that ventilation duct are monitored.
If those exceed 6 certain levels, you can go,through the standby gas treatment 7 system before you discharge to the atack.
8 We are going to talk about the upper dryweJ1, now, 9 which is the region shown here. These engineering drawings are 10 intended to show the platform levels in blue. And we will step 11 through those very quickly. The entry level into the drywell 12 is up here. It contains a personnel lock. It contains stair gg 13 access to the lower levels. And it is really a big flat U
14 working surface. This blue is tended to indicate the grading.
15 And what you see are the tops of the SRVs. You see the tops of 16 the isolation valves. It will be a very clean entry, a lot of 17 room. There is a monorail, two monorails overhead that can 18 pick up any of the valves, any of the HVAC equipment, move it 19 over to the equipment hatch sc it can be taken out of the 20 containment.
j 21 That showed 360 degree access. As you get down lower l
22 in the plant, you can see you are blocked here by the 23 steamlines, but we have platforms that go around the edges. We l
24 give you access to the penetrations, the valves for maintenance 25
(]) inspection services. Here you see the main steamlines going Heritage Reporting Corporation (202) 628-4888 i
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44 kl 1 out.
2 Directly underneath them, we have a layer of 3 equipment that is for, that shows the ECCS piping coming in.
4 Now, this is Division B. This is Division C. And the top half 5 of the drawing is all Division A. And we try and keep all of 6 the Division B equipment, the diesel generators and all the 7 piping and ECCS pumps in this quadrant of the plant. Likewise
_ 8 for Division C and Division A work.
9 You can see the electrical penetrations coming in 10 from each one of the quadrants here. You can see the piping 11 penetrations.
12 MR. REMICK: Is it one of these figures that you were 13 going to show us why the SRV blowdown lines could not be in 14 that space?
15 MR. WADE: There is not any real way -- there is the 16 reactor vessel, support skirt, the haunch on the pedestal and 17 this is the reactor pedestal. There is really no way to get 18 down through the annulus between the reactor vessel and the 19 pedestal down through this area and come out with discharge 20 under the pool.
21 MR. REMICK: Convi.nce me. I hear the words, but.
22 MR. WADE: Well, you've got, I think it's two feet 23 between the reactor vessel and the shield wall. And there is a 24 layer of insulation in here and, so, it is very difficult to
()
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25 get a 10-inch pipe down through this region and have proper Heritage Reporting Corporation (202) 628-4888
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45
-' .$ 1 insulation. And then you would have to go through the skirt.
2 MR. MICHELSON: Why does it have to corae down that I 3 route?
4 MR. WADE: Well, the other routes would be down 5 through the pedestal.
6 MR. MICHELSON: Down through the floor?
7 MR. WADE: Well, it does come down through the floor.
8 MR. MICHELSON: Yes, but you can put concentric 9 piping on after it penetrates the floor. That was the 10 question: Why isn't it concentric? Because I think it used to 11 be. Our discussion about a year and a half or so ago on ABWR, 12 you were using concentric piping.
13 Now, I will admit that the NRC will buy single wall 14 piping because that came through as another whole issue.
15 MR. EBERSOLE: And that gets back to the fact that 16 you can't vent this thing. The fact that you can't vent it 17 means you can't live with a pipe failure. On the other hand, 18 if you could vent it, you could live with a pipe failure.
19 MR. MICHELSON: With the right kind of vent.
20 MR. EBERSOLE: Yes.
- 21 MR. MICHELSON
- But without concentric piping in 22 there, there are a few questions you have to look at closely.
23 MR. EBERSOLE: Why is our pipe so much better than 24 the German pipes? Or what do the Japanese do?
25 MR. WADE:
(]) Do we have a typical Mark II system where Heritage Reporting Corporation i
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46
/~h
- the pipes penetrate the floor and are routed down to the 1
2 quenchers down in this region of the pool. Single wall pipe 3 built to I think it's Class B.
4 MR. EBERSOLE: What can you tell me about the German 5 and Swedish designs? Do they use single wall pipe? Am I ,
6 wrong? ,
7 MR. WADE: The new proposed Swedish design has double 8 walled pipe, but I am not sure that one has been built.
9 MR. EBERSOLE: They make good pipe.
10 MR. MICHELSON: Well, we will get into it later on, 11 but it will be a question that I think will require close 12 attention.
13 MR. EBERSOLE: It seems such a simple thing to lay to 14 rest.
15 MR. MICHELSON: There is no obvious reason why you 16 couldn't use concentric piping. It is not a layout question.
17 MR. WADE: Concentric piping, if you have a lot of 18 bends in there to reduce --
19 MR. MICHELSON: You aren't showing any bends in your 20 design. Well, you are showing a little bit. I can't tell if 21 they are above or below the floors.
22 MR. WADE: There are numerous bends in the pipe 23 between this point, which is an anchor and the quencher which 24 is an anchor to provide necessary flexibility in the line for I
25 thermal expansion.
( })
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1 MR. MICHELSON: And they are below the floor, you're 2 saying?
3 MR. WADE: The quenchers are down here and the 4 penetrations are through the floor.
5 MR. MICHELSON: The bends are below the floor.
6 MR. WADE: The bends are below the floor. There is 7 flexibility built in the pipe between this region here so you 8 can have thermal expansion of that piping.
9 MR. EBERSOLE: I believe there was some problems with 10 the Mark II pipes -- you can correct me if I'm wrong -- in this 11 aspect. That you had to do something special about them.
12 MR. WADE: I think we built them to the quality to 13 Group B.
14 MR. MICHELSON: We will get into this whole thing 15 later in greater depth.
16 MR. WADE: The other major drywell area is what we 17 call the lower drywell. And we have the reactor vessel here.
18 We have the pedestal in this region. We have the reactor 19 internal pumps, 10 of them, in 205 is our fine motion control 20 rod drives going in the bottom.
21 There is greatly improved access to the lower drywell 22 compared to what we have had in any previous plant. There is a 23 tunnel through the suppression chamber here for equipment. You
-24 see in this case, they are bringing in the motor for
() 25 replacement on the RIP.
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l 48 7_
(_) 1 On the other side, 180 degrees away, there is a 2 personnel tunnel with an airlock so you can get in on that 3 region. We have stair access up to this level which is where 4 we do all the servicing of the pumps. And there is another 5 service level up higher where you can get up to the he,at 6 exchangers where we have ladder access. There is not much 7 access up to that level.
8 One thing I wanted to point out on this drawing are 9 the vacuum breakers for the wet well airspace to the drywell, 10 come through the pedestal at this point, so they go from the 11 wet well airspace into the lower drywell.
12 MR. EBERSOLE: I dJn't know whether this is a good 13 place to talk about it or not, but are you providing features 14 that guarantee if you have a meltdown that it will not be a 15 pressurized meltdown and result in discharge into this area 16 with more violent results?
17 I believe there is a fundamental criteria and we 18 shall not have in any reactor a pressurized meltdown. Anybody l
19 can correct me.
20 MR. MICHELSON: We will get to that when we get to 21 the severe accidents.
22 MR. EBERSOLE: But this geometry here is related to 23 that.
24 MR. MICHELSON: We will go into the severe accident 25 scenarios in some depth at a later date. Right now, we are
(])
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49 I'",)
1 just trying to get an overview.
2 Would you point out the reactor, the relief valve 3 tailpipes on that drawing?
4 MR. WADE: They don't show up on this one.
5 MR. MICHELSON: They don't show on this one? .Okay.
6 I see some pipes there, but I wasn't sure what they were.
7 MR. WADE: This is ventilation -- these are 8 ventilation ducts into the reactor area and into the lower 9 drywell and there are some cooling waterlines come down here.
10 And the power to the reactor, internal pump motor, the 11 instrumentation from the fine motion control rod drives goes 12 out the drywell connecting vents.
3 13 MR. MICHELSON: Thank you.
(J 14 MR. WADE: Just a plan view to get you oriented.
15 This is at the pump level and you can see the 10 pumps around 16 here and see the control rod drives. You can see the piping 17 for the hydraulic systems. Again, 360 degree access.
~
DR. KERR: Where can I see the control rod drives?
19 MR. WADE: The control rod drives are here. This is 20 the control rod drive hydraulic piping, those six blocks.
21 DR. KERR: Thank you.
22 MR. WADE: There are 205 drives in there.
23 MR. MICHELSON: Are the drive modules, the control 24 modules in that area, also?
() 25 MR. WADE: No, they are outside in the reactor Heritage Reporting Corporation (202) 628-4888
50 1 building.
2 MR. MICHELSON: All that little piping has to come 3 in?
4 MR. WADE: All the piping comes in the top and bottom 5 of the tunnels.
6 MR. MICHELSON: Top and bottom of which tunnel?
7 MR. WADE: We were looking at this level right here, 8 the control rod drive modules are on the base map. The piping 9 comes up and goes through the top of the tunnels. There is 10 half on this side and half on the other side. It comes up here 11 and then it is routed to the control rod drives.
12 MR. MICHELSON: Okay. I see where they are. Thank 13 you.
14 MR. WADE: I wanted to show this picture which is 15 high in the suppression pool area just to show you that they 16 have the eight vacuum relief valves going through the pedestal 17 which is this region right here from the suppression chamber 18 airspace to the lower drywell. This represents around here a 19 working platform to get to the valves and other equipment in 20 there.
21 This is a pressure suppression type containment where 22 we have blowdown to ambient conditions. We have our 23 suppression pool inside the containment. We use it for LOCA as 24 the quencher for the steam that is released. Also, our SRV and
() 25 their discharge pipes and quenchers are located in there. So, Heritage Reporting Corporation (202) 628-4088
51 1 it is the heat sync for that. Any water that is pumped into 2 the containment from the suppression pool either in sprays or 3 that goes into the reactor vessel eventually runs back down 4 into the pool so we have a recirculating system.
5 The horizontal vent configuration is exactly the same 6 that we used in Mark III. We have 28-inen diameter vents and 7 they are 4.5 feet apart and the same distance off the bottom.
8 So, we have the large data base that is available from the Mark 9 III pressure suppression tests for chugging and CO loads. We 10 are using the same quenchers, so, we have the SRV data base 11 that we had before. .
12 Yes?
13 MR. EBERSOLE: That bullet says, "Simple blowdown O 14 vent system." If it is simple, it is not like the ones you 15 presently use because they involve electrically actuated air 16 operated valves which require environmental resistivity and a 17 host of other things. What have you done --
18 MR. WADE: This is meant to be the horizontal vent 19 system.
20 MR. EBERSOLE: I was thinking about the SRV.
21 MR. WADE: My next slide, I will give you a little 22 picture of this.
23 MR. EBERSOLE: Okay, thank you.
24 MR. WADE: We have the reinforced concrete
{} 25 containment structure that we have talked a little bit about Heritage Reporting Corporation (202) 628-4888
52 a-
/~'S -
(/ 1 before.
2 MR. EBERSOLE: No, is the liner carrying any of the 3 pressure load or is it all on the concrete?
4 MR. WADE: It is all on the concrete.
5 MR. EBERSOLE: And is the liner separated from the 6 concrete?
7 MR. WADE: No. It is attached to concrete.
8 MR. EBERSOLE: Well, is there a gap behind the liner 9 for leak retention --
10 MR. WADE: No. The liner is put in place, at least 11 the wall liner is put in place and is used as a form.
12 MR. EBERSOLE: As a form, okay. So, this is an r' 13 integrai liner.
t 14 MR. WADE: Yes. The floor liner on the botton is 15 wolded on later after the floor is in place. It is welded on 16 inserts.
17 MR. EBERSOLE: How much of a gap on the bottom is 18 thore, then, between the liner and the concrete?
19 MR. WADE: Nominally, nothing.
20 MR. EBERSOLE: It's small. Okay.
21 MR. WADE: I mean there is no purposeful gap under 22 there. We have the primary containment as you saw in the first 23 slide is completely enclosed by a secondary containment. And, 24 of course, the scrubbing that we have, the drywell discharges
() 25 through the suppression pool and any releases that might occur Heritage Reporting Corporation (202) 628-4880 f
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\- 1 from the suppression chamber, airspace are thereby scrubbed.
2 MR. EBERSOLE: How many hours of storage do you have 3 in the suppression pool before you have to start cooling it?
4 MR. WADE: How many hours before we start cooling it?
5 MR. EBERSOLE: Yes, before it overheats and you can't 6 pump anymore.
7 MR. MICHELSON: The norme' assumption is for a 8 30-minute startup on the ECCS systems which provide cooling for 9 the pool.
10 MR. EBERSOLE: That doesn't define the availability 11 of AC power. Does it?
12 MR. SAWYER: Let me take a crack at that.
13 We are in the process right now, I'd say, about a 14 third of the way through cranking all of the mathematics,of 15 ERA. But to get some pre-judgments of how to bend things, of 16 course, we have done some pre-calculations. And it is 17 somewhere in the neighborhood of 8 plus hours is when you've 18 got to start worrying about whether you are pushing the 19 containment.
20 MR. EBERSOLE: I thought it was up pretty high.
21 MR. WADE: That's for the non-LOCA case. Pure 22 isolation.
23 MR. SAWYER: Yes, station blackout. But that 24 includes blowing the plant down to a low pressure as far as
[]) 25 part of the procedures so that if you run out of gas, at least Heritage Reporting Corporation (202) 628-4888
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54 g
! I when you run out of gas, you are at low pressure.
2 MR. EBERSOLE: That may not be the pinch point anyway
- 3 though. Thera may be a hot spot somewhere else. Right? Out
! 4 in the other environment. You might have temperatures risings 5 somewhere else-than in the suppression pool. That is not 6 necessarily the fastest probable.
! 7 MR. SAWYER: That's not necessarily, but I think as 8 we looked at things like control room hee.t up before and this j 9 design is enough similar in that area. I think we will come i 10 out saying that that is the pinch point.
11 MR. EBERSOLE: Okay, thank you.
12 MR. REMICK: Your previous slide said "Reactor 13 blowdown to ambient conditions." What did you mean by that?
br-i 14 You mean it wasn't sub-atmospheric to start? What do you mean i
i 15 by "ambient."
16 MR. WADE: Well, that just means the reactor blows 17 down until it reaches equilibrium with the drywell/ wet well 18 pressure. You know, near relatively low pressure.
19 MR. REMICK: I am not sure I'd call that "ambient" 20 after it blew down. It doesn't matter. Just so I understand 21 what you meant.
22 DR. KERR: As long -as you know what his ambient is.
23 MR. WADE: Here is a picture of the vent system in 24 the ABWR from the upper drywell, the dryuell connecting vents.
() 25 There are about 10 of them, about 1 meters by 2 meters square Heritage Reporting Corporation (202) 628-4888
55 T
("J s l that connect into the lower drywell through an opening here and 2 continue on down in a 1.2 meter diameter pipe. We have the 3 typical horizontal vent system into the suppression pool.
4 I have tried to show on this drawing what our initial j 5 suppression pool level is part way up on the equipment and 6 personnel tunnels. After a LOCA, of course, we have drawdown.
7 We take water out of the suppression pool. We cool the reactor 8 vessel. We fill the steamlines. There is water in transit in 1
9 the spray systems and we can flood the lower drywell right up 10 to this overflow point and then the water spills back. So, 11 after drawdown, we always have at least two feet of coverage 12 over our suppression pool vents --
13 DR. KERR: How do you flood the drywell?
O 14 MR. WADE: Pardon?
15 DR. KERR: How do you flood the drywell?
16 MR. WADE: Well, there is no purposeful means of 17 flooding the drywell. He have said if there is a small break 18 in the bottom, the drain line, the water ca'n come in there.
19 Some of the water might be carried down and get in here. There 20 is no purposeful flooding, but we have accounted for this 21 volume of water in our drawdown calculations.
22 MR. EBERSOLE: Are those vent lines encased in the 23 concrete wall?
24 MR. WADE: Yes. This is a prefabricated steel
(} 25 pedestal. It is made in' modules. And it is filled with Heritage Reporting Corporation (202) 628-4888
56 O 1 concrete after it is put in place.
2 MR. EBERSOLE: Why couldn't the blowdown lines be put 3 in there?
4 MR. WADE: There are 18 of them and there are also 5 some structure recuirements on this pedestal and we just really 6 can't get any more lines down through here and do the welding 7 and stuff like you need to do.
8 MR. EBERSOLE: So, it is too crowded for the blowdown 9 lines.
10 MR. WADE: It is too crowded. i.
11 (Continued on next page.)
12 (Discussion off the record)
() 14 -
15 16 17 18 19
, 20 21 22 23 24
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. (\ 1 1 01. WADE: I think if you look at this picture there 2 are ten events there. You'll see that we have eight places 3 through here. We have vacuum breakers and there's no place --
4 these are all steel diaphragms that go across, but there's no 5 - place now to get down 18 SRV discharge lines.
6 MR. MICHELSON:
I think we're running a little behind 7 schedule. So maybe we better wrap this one up.
8 MR. WADE: Okay. I'm done, except for the last 9 slide, which you can look at in the material, you know. We 10 think we've made improvements. We've used the existing data to 11 come up with a pressure suppression system.
12 I think we've come up with a very strong structure.
13 We've looked at separation and personnel access from the start O 14 of the design. We have local servicing areas. We've looked at 15 radiation levels, tried to get those less than 50 minrem per 16 year, vhich is a very good target, and minimized the piping and 17 electrical in the ouildings.
18 DR. KERR: What is the design basis for your 19 containment? Is it the 45 pounds gauge?
20 'MR. WADE: No. I mean, you arrive at the 45 pounds 21 gauge after having made some performance specifications, and I 22 just wanted --
23 DR. KERR: what I'm asking is, is it the 3 WASH-1400 24 era to which we are designing containment or has there been 25 some change in design based on the fact that we at least now
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('>1
'- 1 recognize the possibility of severe accidents?
2 MR. WADE: Well, --
3 DR. KERR: I assume the 45 PSIG c'ame from an analysis 4 of the design basis accident that has been in use since '71 or 5 whatever. Is that the case?
6 MR. WADE: Yes. You know, we did realize that there 7 are severe accidents.
8 DR. KERR: I'm trying to understand whether this was 9 factored into the design of your containment.
10 MR. WADE: Yes, it is. For instance, in one case it 11 doesn't show, it doesn't show very well on this slide, but the 12 containment liner goes across the bottom of the suppression
, 13 pool, across the bottom of the pedestal and through this area.
\- 14 And wa have put additional protection between the containment 15 liner and the bottom there so that any debris that would fall 16 down here from a severe accident would have to go through 1.6 17 meters of containment of concrete before it reaches the 18 containment liner.
19 DR. KERR: Okay. So that's one feature. Anything 20 else?
21 MR. SAWYER: This is Craig Sawyer again. We have 22 also committed to meet the factor load 100 percent metal water 23 reaction edition in terms of overpressure. You were worried 24 about where the 45 pounds came from.
(} 25 DR. KERR: You anticipate if you do have core melt Heritage Reporting Corporation (202) 628-4888 e
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1 that the volume below'the reactor will have water in it or not?
2 Was that taken into account in any fashion in your design?
3 MR. SAWYER: Well, we are just in the process of 4 doing those. kind of calculations right now.
5 DR. KERR: Okay.
6 MR. SAWYER: You know, as Mr. Wade mentioned, for one 7 thing, you get-the water there automatically.
8 DR. KERR: Since your containment is already 9 designed, it's fair to say those calculations haven't 10 influenced this design?
11 MR. SAWYER: There is one happy future about this 12 design. In terms of looking at this design, with the severe 13 accident that happened a coupla of years ago, that I'm real O 14 happy about. Which is, it is the same as the MARC-III. And 15 that is that if you have LOCAs you have water underneath there 16 for the core to begin with. If you don't have LOCA and you 17 start with a dry dry well, you get lateral attack on the 18 pedestal wall and a reasonably rapid -- by reasonably rapid 19 this probably means a couple of hours. Because you have those 20 ten locations where the vents are, you have a minimum thickness 21 of concrete there and you will end up getting the suppression 22 pool water to communicate with the dry well by the nature of 23 the core melt event itself.
24 And of course, if you go through those kind of
(} 25 considerations, you worry about recovery as we will. In a Heritage Reporting Corporation (202) 628-4888
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! (ss) 1 recovery scenario that follows a core melt event will end up 2 with water in the dry well flowing through the same hole the 3 core melt came through.
4 DR. KERR I'm a great believer in serendipity, but 5 it does surprise me a bit that there has been this little 6 change in -- this is virtually the containment design that you
-7 would come up with if you used the old design basis accident 8 with possibly the one exception of hydrogen.
9 And I guess we were just lucky enough that we 10 originally designed these containers to take care of severe 11 accidents even though we didn't believe they would occur.
12 That's sort of the only conclusion I can reach.
13 For example, is there equipment that you would need O 14 to work within the containment? And this I guess we'll get to.
15 So I'll reserve that question until later.
16 MR. MICHELSON: Yes. We're going to get to much more 17 detail on all the severe accidents.
18 DR. KERR But it does appear that severe accidents 19 had very little to do with this containment design. Which 20 means, I guess, that they had been lucky enough that the MARK-21 III design was almost what one woulo want if one had designed 22 with severe accidents in mind.
23 MR. MICHELSON: I believe we'll have to proceed. But 24 since we show a break at 10:00 O' clock, I guess it would be
{} 25 best if you wanted to wait until right after the break? How Heritage Reporting Corporation (202) 628-4888
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} 1- long do you'think you need, do you know? Maybe it would be
. 2 best now to break first?
3 104. CORK: There is one more. item on the agenda 4 before the NRC-ar.' it is a very brief item so I would propose 5 that we could do that'now before the break.
6 MR. MICHELSON: All right. While the next speaker is 7 . coming up, maybe I could ask GE. I had asked the staff, but 8 maybe I'll just go right to the source.
-9 Some of these cutaway colored drawings are extremely 10 helpful in following discussions throughout these meetings.
11 This type of drawing. I think you actually, I've seen a couple 12 of other cutaways of this in color.
13 'Could you provide us a bunch of colored drawings of O 14 these plants? By a bunch I mean maybe about 15 copies, of this 15 cutaway and any other good cutaways that you have, and then the i
16 members I'll be sure will have at least one. If they don't 17 bring them to the meeting, it's their problem, but I'd like to 18 make sure they all have them to follow.
19- Because it's hard to see a cutaway this far back and l
20 yet they are extremely helpful in getting a perspective on the
- 21 plant. At least I find them so.
I 22 MR. SAWYER: I'll be happy to do that.
! 23 MR. MICHELSON: Just send them to Richard and he'll l 24 take care of it.
25 (Slides shown) l l
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() 1 MR. CORK: My name is Joe Cork. I'm the Program 2 Manager for the ABWR Certification Program. I am also the 3 Program Manager for the GE participation in the EPRI 4 Requirements Program. And as Dr. Wilkens mentioned, the close 5 paralleling of the EPRI requirements and the ABWR certification 6 program are important and headed at GE by the same individual, 7 myself.
8 The item on the agenda that I will briefly discuss 9 next is the discussion of the new balance of plant scope. I 10 think I can deal with this subject in two slides.
1.1 This slide is intended to show the scope of the 12 existing licensing submittal. The scope of our licensing 13 submittal is referred to as the nuclear island scope. And that 14 is defined here as containing equipment within the reactor 15 building and the control building.
16 So al? these systems, the structures and the 17 components within the reactor building, including the nuclear 18 steam supply, the primary and the secondary containments, the 19 ECCS systems, the RHR systems, the diesel generators, the fuel 20 handling equipment, all these systems and equipment that are 21 essential to safety, housed in the reactor building, are within 22 the licensing scope.
23 So also in the control building, the control room 24 complex and the other features within tne control building are 25 part of our licensing submittal. '
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() 1 So we refer to this as the nuclear island scope.
2 As we have progressed with our certification program 3 and nave met with the ACRS subcommittee and full committees as 4 well as the Commissioners, the feedback we have received is 5 that in order for certification to work.as it is intended, 6 plants that are certified should consist of mos't if not all of 7 the plant.
8 And so Chairman Zech has encouraged GE in the three 9 meetings that we've had with him to add to our nuclear island 10 scope major portions that are outside that scope, such as the 11 turbine island and the rad. waste facility.
12 MR. MICHELSON: Does the turbine island include the 13 cooling towers?'
14 MR. CORK: No, it does not. I have a chart that shows 15 you what is in and what is out.
16 But the turbine island would include, of course, the 17 turbine building, the turbine itself and the generator, the 18 feedwater systems, the condensate systems, basically the power 19 side of the plant, and of course the rad. waste facility. And 20 this view was shared not just by Chairman Zech but by the ACRS 21 members themselves as well as the House Authorization Committee 22 markup of the '88 budget had some guidance in their 23 authorization to encourage certification applications to 24 consider the total plan.
25 So with this type of encouragement, and working with
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() 1 DOE, GE is prepared to expand their licensing scope in the near 2 future to include more than the nuclear island, to include both 3 the turbine island and the rad. waste facility.
4 And the next chart I attempt to show the scope of our 5 too soon to be amended SAR. Let me move this up here.
6 Number two is the reactor building. We identify the 7 equipment features in there.
8 And Number three is the control building.
9 And so this block I've referred to as the nuclear 10 island. We will be adding the turbine island and the rad.
11 waste facility, leaving the items that are noted in green here, 12 which is the service building the switchyard, the cooling tower 13 and the ultimate heat sink. It is not our intent to exclude 14 these from our scope because we think that all the important 15 safety parameters associated with those features will be 16 identified and spelled out clearly in our safety analysis ,
17 report.
18 MR. EBERSOLE: Joe?
19 MR. CORK: Yes.
20 MR. EBERSOLE: Item 9 in particular is the heart of 21 the safety problem.
22 MR. CORK: The ultimate heat sink.
23 MR. EBERSOLE: I'm always astonished to see Item 9 24 not included as within your safety scope.
25 MR. CORK: Well, Jesse, your point is valid. It is
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\_/ 1 very essential to safety. It rejects the heat and of course 2 therefore it is important. But it is very site dependent, 3 whether that's a cooling pond or a cooling tower or an ocean, 4 you know --
5 MR. EBERSOLE: Even so.
6 MR. MICHELSON: We will become curious later on as to 7 how you do a PRA without Number 9 in it.
8 MR. CORK: We will identify on our PRA the 9 assumptions for the ultimate heat sink and ensure that those 10 assumptions are followed through and dealt with by the 11 applicant.
12 MR. MICHELSON: Yes, there will be required, highly 13 detailed interface requirements, in order to assure those (E) 14 assumptions are met.
15 MR. CORK: I would like to conclude by saying that I 16 said that GE and DOE are tentatively planning to augment this 17 existing nuclear island scope to the total plant. We as of yet 18 have not done that.
19 We expect to do that and we would seek from the 20 Cotamission their endorsement and resolve to review these 21 modules on the existing schedule that has been established.
22 MR. MICHELSON: Would you anticipate at least 23 designing what I'd call a typical or prototypical ultimate heat 24 sink arrangement recognizing that on a case by case basis it
{} 25 would change? Then I would believe your PRA a little more.
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n 1 MR. CORK: I don't know how you design a typical 2 ultimate heat slak because it's very dependent on the site 3 itself and whether it's one of the three that I just discussed.
4 MR. EBERSOLE: The only thing you could do is make it 5 operate from there.
6 MR. CORK: Are we talking about ultimate heat sink?
7 Oh, towers?
8 MR. MICHELSON: We'll look forward to looking at how 9 you handle that one.
10 MR. NYLIE: In expanding into including the balance 11 of plant, are you going to then produce specifications for the l 12 components?
l 13 MR. CORK: Yes, we are. We will specify to the O 14 procurament level a detail that Dr. Wilkens talked about, the l
l 15 information --
l l 16 MR. WYLIE: For all components?
i 17 MR. CORK: Yes.
18 MR. MICHELSON: As a part of the certification 19 process, do you anticipate that the specifications are included 20 in the certification?
21 In other words, to put it more aimply, what do you 4
22 think is going to be certified?
23 MR. CORK: Well, we say that inside the cross-24 hatched areas, we will specify systems, equipment and structure
{} 25 and the performance capability of equipment, and the Heritage Reporting Corporation (202) 628-4888
67 I) 1 configuration and layout of the areas, and the orientatior,of 2 equipment relative to each other, down to a level of
-3 procurement.
4 MR. MICHELSON: And do you think down to that level 5 it will be certified? In other words, is that the level at 6 which it is frozen, so to speak?
7 MR. CORK: Yes.
8 MR. MICHELSON: Presumably in this process we are 9 going to freeze everything.
10 MR. CORK: The final design to be unchanged.
11 MR. MICHELSON: So the specification then is a part 12 of the freeze process. Is that what you think?
13 MR. CORK: That is how we view it.
k- 14 MR. MICHELSON: Thank you.
15 MR. WYLIE: That is a procurement spec. You could 16 send it our for bid.
17 MR. CORK: That's right.
18 MR. EBERSOLE: There are, and there have been, and I 19 guess there are now, plants in which the ultimate heat sink is 20 in fact the atmosphere with force convection. It might be 21 worthwhile thinking about that as a standard feature, because 22 it eliminates all this variability about the creeks and rivers 23 and wells and whatever else.
24 After all, we'ra sitting on a sea of air. It's hard 25 to get away from it. And it might be worth thinking about.
(T
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s,_/ 1 MR. MICHELSON: I think we'll have to proceed since 2 we're running out of time. We'll have a break now, if it's all 3 right, and then we'll pick up with you right after the break.
4 (Whereupon, a brief recess was taken.)
. 5 MR. MICHELSON: Let's proceed. We're running a 6 little bit late. And I believe Dino is going to tell us a 7 little about the Japanese trip next. So proceed.
8 (Slides shown) 9 'MR. SCALETTI: Good morning. My name is Dino 10 Scaletti. I am the NRC Project Manager for the Advanced Boiling 11 Water Reactor. I am here to discuss our recent or February 12 trip to Japan.
13 With me I have Mr. Lester Rubenstein, our Assistant 14 Director for Special Projects in Region IV Plants; Mr. Charlie 15 Tinkler who is from the Plant Systems Branch of NRR and I am 16 with the Standardization and non-Power Reactor Project 17 Director. Sometimes we get confused.
18 I am pleased to be here with you this afternoon or 19 this morning to be able to discuss some of the highlights of 20 our trip to Japan. We ended up with having five very full days 21 of meetings the week of February 2nd in Tokyo and outside of 22 Tokyo.
23 I will take a few minutes to discuss some of these 24 things.
[} 25 I'd like to structure this presentation along the Heritage Reporting Corporation (202) 628-4888
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() 1 lines of purpose of our visit, visit objectives, organizations 2 we met with, and then at the end, questions and responses, 3 throw the meeting open to any questions you may have. The 4 questions and responses here are the ones that we had prepared 5 and discussed with the Japanese.
6 I would like to state that we have prepared a trip 7 report to cover the five days of meetings we had in Tokyo so 8 this is available for anybody to review and to look at.
9 I believe the ACRS has copies of that trip report.
10 Contrary to what some people may state, the purpose 11 of this trip is to support the design certification review for 12 the advanced boiling water reactor. I have been accused of 13 many other things, but this is the purpose.
14 Now, one of the reasons for this is that as you are 15 well aware, the ABWR is being designed by three designers, two 16 of them which are Japanese -- Toshiba and Hitachi -- although 17 the application for certification in the U.S. is being 18 supported solely by General Electric Company.
19 The staff felt that it would help us probably to gain 20 some insights into the Japanese design process and the Japanese 21 regulatory process.
22 MR. MICHELSON: What part does DOE play in this if GE 23 is sponsoring this totally?
24 MR. SCALETTI: DOE provides funding I believe. I'll
{} 25 let General Electric address that, if they would.
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() 1 MR. CORK: Yes. c DOE sponsors the certif1'ation, so 2 they are paying for General Electric to translate the Japanese 3 APWR design.
4 MR. MICHELSON: So it's not -- I think Dino said it 5 was a GE show. You are not the sole sponsor.
6 MR. CORK: We're funded by the Department of Energy.
7 MR. MICHELSON: You're funded by DOE. Okay. Thank 8 you.
9 MR. SCALETTI: The application for certification says 10 General Electric, not Department of Energy.
11 MR. RUBENSTEIN: Excuse me. Lester. Rubenstein.
12 That's an important point. Because the application in the 13 United States, particularly in regard to things like quality 14 assurance, has to be dealt with by -- they have to demonstrate 15 to us that they have the control over the quality of the design 16 itself, and when they talk to the Commission, they talk with 17 authority, because they've had processing procedures in place 18 to allow them to participate in the design features.
19 MR. MICHELSON: Thank you.
20 MR. SCALETTI: Our visit objectives were to establish 21 contact with appropriate Jartnese organizations that were 22 involved in the ABWR licensing process and licensing 23 construction and design. .
24 We started by establishing contact with many 25 representatives who are here at the U.S. NRC, Mr. Togo, with
)
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( ). 1 TEPCO, Tokyo Electric Power Company and Mr. Omoto, who has been 2 very helpful to us, and also the Japan Electric Power j 3 information center here in Washington, D.C., and Mr. Hatano, 4 who were all very helpful in establishing, helping us set up an 5 agenda and contacting the appropriate people to meet with in 6 Japan and also through the Department of State to -- our 7 contact through the Department of State and for officially 8 setting the meeting up.
9 The organizations I'll discuss shortly in the next 10 slide I believe it is. But one of the reasons for going again 11 was to compare the U.S. and Japanese regulatory requirements.
12 It was just for us to gain an understanding of the Japanese 13 regulatory requirements and the regulatory process. As this -
14 design is supposed to be international, the design as being 15 built in Japan for the most part I understand is being carried 16 forth into the U.S. design certifications. Clearly, much of it 17 was designed to Japanese requirements.
18 MR. MICHELSON: What are the requirements that were 19 identified? Is there some kind of a standard review plan or 20 anything like that?
21 MR. SCALETTI: I am not aware of a Japanese standard 22 review plan.
23 MR. MICHELSON: You also talk about standardizativa 24 policy. Have they issued a standardization policy statement as 25 such?
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1 72 1 MR. RUBENSTEIN: No, they have not. But in some of 2 their documents they, talk about four phases of light water 3 reactors, the upgrading, they talk about four phanes of 4 upgrading. And they had a sophistication committee for the 5 improvement of light water. reactors that we've dealt with over 6 here.
7 And if we talk about BWRs, they would say that they 8 took the best of Western technology and put that in as their 9 base case.
10 MR. MICHELSON: I was mostly interested in 11 documentation that one might be able to even read.
12 MR. RUBENSTEIN: There is some documentation.
13 MR. MICHELSON: But there is not a policy statement l
/~T V 14 as such?
15 MR. RUBENSTEIN: No. Their standardization is really 16 more, they plan to evolve over four different periods in BWR or 17 the light water reactors.
18 2'R. MICHELSON: this is mostly in people's heads, 19 then?
20 MR. RUBENSTEIN: No, no. It's written out. For 21 example, I think that the second or third page, they put 22 together a survey mission on LWR technology, for a 23 sophistication steering committee.
24 And in that framework, they came and they started to
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(_ 1 are you going to get a little more out of the reactor's 2 performance. ,
1 3 So in that period is one of the stages of upgradii.g.
4 Then later on they got to the ne.it two stages where 5 there is the 1,000 megawatt gas design and then a fourth is a 6 as yet hazy to us stage where they will probably go to 7 something equivalent to our advanced plant.
8 MR. WYLIE: But that's development.
9 MR. RUBENSTEIN: Yes.
10 MR. WYLIE: But what about regulation? How do they 11' control that process?
12 MR. SCALETTI: The Japanese clearly do not have a 13 process akin to our certification process.
U 14 MR. WYLIE: So they don't really control it?
15 MR. RUBENSTEIN: No, they do. Let me answer that.
16 What you have to do is not think of MITI as wearing 17 two hats. You have to think of MITI as having both the 18 development and the regulatory hat on. And specifically in the 19 case of the ABWH, we met with the appropriate people in MITI 20 and the examiners or the reviewer of the ABWR had been involved 21 early on in the early design decisions. So it is truly a 22 collegial thing where TEPCO played a very large role in saying 23 what the requirements would be. GE, Hitachi and Toshiba played 24 a large role in the design. And MITI was as well involved in 25 many of the design requirement decisions. I don't know how
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(-) 1 much they got into directly in the development stage.
2 GE could probably elaborate more in detail on how it 3 went on over the past four or five years. But I want to leave 4 just the impression that MITI has n.uch more than a regulatory 5 role.
6 MR. WYLIE: I still don't have any idea about 7 control.
8 MR. SCALETTI: The standardization policy per se is 9 more akin to our, perhaps our replicate plant policy or 10 duplicate plant policy. The development of the ABWR, there's 11 two units going in. You can conceivably consider those 12 duplicate plants. They will have a total design. The plant, 13 when they file for their reactor establishment permit, this 14 reactor establishment permit is good for those two units at the 15 Kashiwazaki-Kariwa site.
16 Now, for another ABWR to be constructed, another 17 reactor establishment permit at'another site would have to be 18 granted, an application filed and the permit granted. They do l
19 rely heavily upon the review that has gone on in the past so l
l 20 they do give credit as we would for a replicate plant or for a 1
21 design certification, the credit is there, although the process 22 of licensing is individual.
23 MR. RUBENSTEIN: Specifically on the standardization 24 policy, they don't have a directly comparable standardization 25 policy because they don't have our body of regulations or our O(~N Heritage Reporting Corporation (202) 628-4888 i
I ' .
I 75 (m_) 1 own internal process of going through the hearings in the same '
2 way we do, 3 When we talk about standardization process, what 4 comes to mind is the change in 10 CFR 52, which you may have 5 heard yesterday. The early site permit, the design 6 certification process gets involved in how we can facilitate 7 and encourage industry to do things by deferring fees, how we 8 can combine the processea of CP and OL.
9 What it really is is a facilitation of the process 10 of licensing, and they're not, in their approach, encumbered in 11 the same way that we are with our institution. They have their 12 own kind.
13 MR. EBERSOLE: As to TMI-2 when NRC invoked the 14 concept of evacuation as critical to the overall safety of the 15 plants, my understanding was that this was bitterly opposed by 16 nearly all foreign countries, because they're crowded. And you 17 can see the results of having invoked that, of course. They're 18 clearly here today.
19 I don't understand what, for instance, the Japanese 20 do. All of these discussions about the quality of plant design 21 safety eventually converge on just the severe accident in the 22 context of personal safety.
23 What do the Japanese do about the matter of the 24 evacuation requirements, if anything? Or sheltering, if they Or do they even think about severe accidents?
{} 25 use that? I Heritage Reporting Corporation (202) 628-4888
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% . 1 don't know.
2 MR. RUBENSTEIN: I don't have a direct answer on 3 sheltering. But every time we address severe accidents, the 4 answer we get is accident prevention. And that is truly their '
5 answer. They work on prevention.
- 6 Subsequent discussions then get into well, we do the 7 PRAs, we do join in the international community in looking at 8 some of these phenomena specifically, and we study severe 9 accidents. And that's how we had to leave it.
10 MR. EBERSOLE: See, you haven't said anything of 11 substance. But the substance here is :errible.
12 MR. RUBENSTEIN: But the answer directly is, we 13 believe in prevention. That's the substance.
) 14 MR. EBERSOLE: So you're telling me there is no 15 consideration of evacuation needs or sheltering at all. Is 16 that true of France and Germany?
17 MR. RUBENSTEIN: I can't answer that for France and 18 Germany.
19 MR. EBERSOLE: So we are unique in the world, we've 20 got to run. l 21 MR. SCALETTI: Well, clearly the Japanese, we were 22 told many times that they don't have to consider severe 23 accidents, although they do consider to a degree. They have 24 done some PRA work on the ABWR. But it is also reflected, the
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k/ 1 ABWR is going to be built on will eventually have, when they're 2 finished, over 8,000 megawatts electric on this one site. So 3 this is -- the concern clearly is not there. And they have 4 stated to us many times in our discussions that for one thing, 5 their diesels don't fail. You had mentioned that earlier on.
6 MR. EBERSOLE: We have a societal barrier to nuclear 7 plants.
8 MR. MICHELSON: We will later have to come to grips 9 with how we're going to consider the severe accident policy on 10 this plant. And I think we'll have to move on for now.
11 MR. SCALETTI: One of our objectives also was to 12 visit some of the component testing laboratories in Japan to 13 view some of the ongoing efforts for the ABWR.
O 14 We had the opportunity while we were there to visit 15 NUPECs Katsuta laboratory and also the Isogo laboratory where 16 we had the opportunity to view at Katsuta the Hitachi-designed 17 reactor internal pumps and we had the fortunate opportunity to 18 see them in operation, to feel them, touch them, listen to l
19 them, and observe some of the test data.
l 20 At Toshiba, they were testing a version of the German 21 pumps. And the same situation, they were running, and we had 22 the opportunity to observe the tests and to have physical 23 contact with the pumps as well as many other comoonent testing 24 situations.
O V
25 The organizations we met with, I'll just quickly go Heritage Reporting Corporation (202) 628-4888 e
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\_) 1 over, was the first day we met with MITI. We had a very 2 information meeting with MITI that lasted most of the day and 3 into the early evening.
4 We established MITI-NRC points of contact between 5 myself and a Mr. -- the name slips me at the moment --
6 Nishiwaki, who I have agreed to send him information as it 7 comes to us and they have agreed to provide us with pertinent 8 information that they received with the understanding that they 9 do not plan on translating everything, and we agree. However, 10 if it is important, I believe they would translate it for us.
11 So we have been sending information to Mr. Nishiwaki 12 at MITI.
13 DR. KERR: I was hoping you were going to tell us you 14 were learning Japanese so you could do the translation.
15 MR. SCALETTI: No. I have a couple other languages I 16 have to learn first.
17 The other organizations we met with, Tokyo Electric 18 Power Company. We met with Tokyo Electric, Hitachi and 19 Toshiba, for the remaining four days in Tokyo and outside of 20 Tokyo, so we spent a little better than a day apiece with each 21 one of the two designers and Tokyo Electric. We were fortunate 22 to be able to go out to Lhe Kashiwazaki-Kariwa site on the 23 Japan Sea side of Japan to visit the actual site of the two 24 ABWR units.
We were able to observe Kashiwazaki-Kariwa Unit 5
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79 1 under construction and we were, the construction was proceeding 2 on schedule. We were quite impressad with the construction 3 process.
l 4 I guess the construction time is estimated to be 78 5 months. They were on schedule and they were maintaining that 6 schedule with one shift.
7 We visited an operating unit at that same site, 8 Kashiwazaki-Kariwa Unit 1, that had been on line for at least 9 three years, two years maybe. It was down for refueling, into 10 the very early stages of its refueling cycle. And again, we 11 were extremely impressed with the procedures from the operating 12 reactor standpoint, the cleanliness of the plant, the 13 maintenance procedures that went on, the refsteling procedures.
14 MR. MICHELSON: Excuse me. Was that a boiling water 15 reactor?
16 MR. SCALETTI: Yes. BWR-5, 17 MR. MICHELSON: BWR-5?
18 MR. SCALETTI: They were all BWR-5s except for the l 19 two ABWR units, I believe.
l l 20 MR. MICHELSON: What kind of containment was this?
l l
21 MR. SCALETTI: MARK-II.
22 The Japanese are very concerned about exposure. A 23 great deal of remote handling on all their operating reactors 24 is reflected in the ABWR also.
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( ,) 1 the plant and it is pretty much automated. And they had three 2 people on the refueling floor who were overseeing the 3 operation. It was operating smoothly with those three people, 4 and quita impressively.
5 MR. SCALETTI: Also, on the turbine deck they were 6 dismantling the turbine for inspection purposes. When they go 7 into a 90-day refueling outage, they seem to go at it with a 8 great deal of zeal and get all their maintenance taken care of 9 in that one time.
10 The days that we had again met with Hitachi we had 11 visited NUPEC laboratory where they were testing the Hitachi 12 pump so we also had the opportunity to go to their site where 13 they were constructing the turbine for the Kashiwazaki-Kariwa
\l 14 Unit 5. We were very tmpressed with the procedures, by the 15 workmanship that was going on in constructing this turbine.
16 They seem to take a great deal of pride in their workmanship 17 and clearly the result of their workmanship.
18 The last day there we met with Toshiba in Yokahama, 19 had an opportunity to visit the Isogo laboratory. Part of it 20 we did not get in to see was the NUPEC facility where they were 21 testing, doing some of the testing for I believe Toshiba. We 22 did look at the testing of the internal pumps again. We were 23 able to review, and to look at the fine motion control rod 24 assembly, the testing and handling equipment for that, handling A
(- 25 equipment for the in*ernal reactor pumps.
(_T)
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(/ 1 We were fortunate to go to one of their advanced 2 simulators. This was not the simulator for the ABWR. I don't 3 believe that that has, the control room has been decided yet.
4 Maybe it has, but I haven't heard.
5 We were clearly impressed with the simulator.
It 6 utilized artificial intelligence, it could be operated by one 7 man sitting down. Whenever there was a transient of any sort 8 it gave some sort of procedures, hierarchical type of actions 9 to be taken by the operator to control the whatever transient 10 it was. So, extremely impressed with the organizations, the 11 designers.
12 MR. RUBENSTEIN: Are you going to mention the one-13 fifth scale?
14 MR. SCALETTI: Oh, yes. Also, we had the opportunity 15 to observe reactor vessel and one-fifth scale with ten scale 16 down and internal pumps operating. And they ran it for us so 17 that we could observe the flow patterns with one pump out of 18 operation and with two pumps out of operation. So you could, 19 the way the setup was, they had used a flat laser beam to take 20 a slice down through the vessel right over two of the reactor 21 pumps so you could clearly see the patterns, the flow patterns 22 with the pumps out of service.
23 MR. MICHELSON: These were operating at full 24 temperature and pressure?
{} 25 MR. SCALETTI: Not temperature, I believe. I don't Heritage Reporting Corporation (202) 628-4888
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s- 1 know about pressure.
2 MR. RUBENSTEIN: It was simulated.
3 MR. SCALETTI: Simulated. Yes.
l 4 MR. RUBENSTEIN: They had a peculiar readout and you !
5 can see the velocity of growth of each of the pumps and these 6 were bar charts, and when I asked them to cut off one pump, 7 both visually with the laser beam and numerically on the 8 screen, one can see the changer flow as the velocity changed, 9 the d'.rection changed, and we watched the reverse flow and then 10 the measured velocity of reverse flow from the pump that 11 presumably had failed was measurable. And then they do that 12 for two, and we were quite able to see that. It was very 13 persuasive.
O 14 M3. MICHELSON: They mocked up the core and you 15 watched the core flow? The core flows through the pumps?
16 KR. RUBENSTEIN: Yes.
17 MR. MICHELSON: You could see the effect on core 18 flow?
19 MR. BUBENSTEIN: Yes, we could see it. It was very 20 persuasive. ;
e 21 MR. SCALETTI: Toshiba had done some work also on the 22 containment. They built a one-tenth scale containment. They ;
23 destroyed it by over-pressure and also I guess they destroyed l 24 it also by seismically-induced vibrations. So there has been 25 an extreme amount of testing with the containment pumps, pretty
(]}
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(_/ 1 much the whole plant.
2 Before we went on the trip, we did a lot of work in 3 the beginning. We initially set forth, established some areas 4 of interest that we would have as NRC staff to go over there 5 and talk with the Japanese. We put together three or four 6 pages of topic areas. Wo sent this out to the review staff, 7 the various review divisions within NRR and asked them to look 8 at this and comment and provide more specific questions that 9 they may have relating to the advanced boiling water reactor 10 and the Japanese regulatory process.
11 That was done. We got back, if you've looked at the 12 report, many, many questions more than we originally started 13 with. And we transmitted these to the Japanese prior to our
(~'s V 14 visit.
15 We were fortunate enough to have them prepare many of 16 the questions, written reeponses to many of the questions which 17 we had before us which we could discuss with them during the 18 meetings with each one of these organizations.
19 So it made our job much easier, our information 20 gathering much easier, clearly.
21 Some of the areas of our concerns were obviously 22 severe accidents, which is a. big concern in the U.S. and we 23 were faced with this each time we met with each of the 24 organizations that they do not have severe accident 25 requirements in Japan.
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84 l s) 1 MR. RUBENSTEIN: Let me give you an example of that.
2 One of the questions that we asked was the standard 3 designs preaently being reviewed in the United States may be 4 required to meet a conditional containment probability of less 5 than one in ten weighed over ten core damage sequences. These' 6 are discussed from the Japanese regulatory perspective. And my 7 notes show one, we have .to regulatory requirements at this time 8 for severe accidents. Two, the probability of severe core 9 damage is very low. We don't believe there is a need for 10 urgent measures. Three, we are studying this. There is a 11 severe accident policy committee that is specifically studying 12 this. And they pointed out some of the people on the 13 committee.
14 And as regards the containments, we are trying to 15 address that from our emergency operating proceuures and that 16 is also on the study. But that is the way we would approach 17 it. And the bottom line is, one more time, we believe in 18 prevention.
19 MR. EBERSOLE: In the prevention category, do they 20 use venting as an escape route?
21 (Conference off the record.
22 MR. MICHELSON: What was the answer?
23 MR. TINKLER: Yes. We had a number of discussions on 24 this with Tokyo Electric Power and we were informed that they have symptom-oriented procedures which would direct the
(} 25 Heritage Reporting Corporation (202) 628-4888
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l 85 ex :
k-) 1 operator to vent at 80 PSIG.
1 2 MR. EBERSOLE: He would get that by deliberate l
3 blowdown?
4 MR. TINKLER: He would take action to open a vent 5 path.
6 MR. EBERSOLE: I mean, what would he do with the 7 primary loop? Would he depressurize it?
8 MR. TINKLER: I can't say conclusively that they 9 follow all of the current EOPs, you know, or the EPGs rev. 4.
10 I don't know if they ascribe to the latest version.
11 Most of the information we received led us to believe 12 that they followed fairly closely the EPGs that have been 13 developed for the BWRs, in which case they would have
()
14 depressurized prior to venting.
15 MR. MICHELSON: But they did have a vent path capable 16 of 80-pound operation?
17 MR. TINKLER: They made it very clear that they l
l 18 thought it was important to have a hardened vent path, pressure 19 bearing vent path.
20 I pursued this, because they indicated that they 21 vented through the standby gas treatment system. And I pursued 22 this as to how they vented through the filters with the standby 23 gas treatment system.
24 I can't say that the response was sufficiently clear 25 to me that I understood it.
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'- 1 MR. MICHELSON: But that was their intent?
2 MR. TINKLER: But that was their intent. But they 3 didn't make it clear -- and MITI made this point clear, that 4 they thought it was very important to have that escape mode.
5 MR. MICHELSON: In the review process for the ABWR 6 there is some question then about whether GE is providing such 7 a vent path and why not, if the Japanese in the past have 8 depended on such vent paths.
9 MR. TINKLER: Well, at least they clearly have 10 provisions for that sort of action now.
11 MR. EBERSOLE: Well, that gets to be complicated.
12 MR. MICHELSON: We'll clear that up later. I don't 13 think we're going to have time, today, Jesse. But when we 14 discuss --
15 MR. EBERSOLE: Along with ventirig, goes makeup.
16 MR. MICHELSON: Yes, right. But when we get to 17 severe accident we're going to pursue with GE the other side of 18 the coin.
19 MR. EBERSOLE: Well, before you get to severa 20 accident --
21 MR. MICHELSON: I understand.
22 MR. EBERSOLE: -- they vent it to what?
23 MR. TINKLER: I understood they vent it through the
' 24 standby gas treatment system, ultimately released to the stack.
25 But when first asked about this, they suggested that the first
({)
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87 ka) action the operator might take would be to proceed to, I guess 2 what you'd' call event oriented procedures.
3 When we asked them about that they said they also 4 have symptom oriented procedures.
5 But we got the impression that the first book they 6 turned to would be event oriented procedures.
7 MR. EBERSOLE: You have a lot of diplomatic 8 ambiguity.
9 MR. MICHELSON: Well, at any rate, we will certainly 10 want to get the story. I'm sure GE will have the story directly 11 from Japan as well, and we'll find out what's happening there.
12 Why don't you proceed, Dino, because we really don't have time
, 13 to explore that in any depth now.
14 MR. SCALETTI: PRAs, we asked many questions on PRAs.
15 And again, the response was no requirements, although TEPCO has 16 done some preliminary PRA work for the ABWR, Level 1 PRA.1 The 17 results are included in the trip report. Also had concerns 18 over fire protection, the interest there. Discussions with the 19 Japanese indicated that they did use physical barriers for 20 separation of safety trains. They also used hardware for 21 detection and suppression, and they elaborated on that quite a 22 bit.
23 But physical barriers was also --
24 MR. MICHELSON: Did they use automatic fire 25 protection?
(])
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(/ 1 MR. SCALETTI: Automatic and manual.
2 MR. MICHELSON: And do they seismically qualify their 3 fire protection?
4 MR. SCALETTI: That I can't answer.
5 MR. MICHELSON: In their PRAs, do they purport to 6 include the external events when doing PRAs, or did they 7 elaborate?
G MR. RUBENSTEIN: Excuse me. And I'm doing this from 9 memory.
10 I think they really did it pretty much for their 11 site. Maybe GE could answer it.
12 MR. MICHELSON: At least GE is going to include it on 13 the ABWR.
O 14 MR. SAWYER: We're going to include it. I don't 15 actually know whether or not PRA studi'es that were done in 16 Japan were included or not.
17 MR. RUBENSTdIN: What they did in the PRA was to do a 18 PRA based on, was it --
19 MR. SCALETTI: Limerick.
20 MR. RUBENSTEIN: Limerich, pardon me. Limerick, with 21 which they use their own failure and even factor numbers for 22 it. So they took that and converted it to their own 23 experience. And we have those numbers and they are available 24 to you.
./~T 25 MR. WILKENS: Let me add one comment on the' l (/
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(_) 1 discussion of the vent issue that we just had a minute ago.
2 The view that the current Japanese BRWs have ADPSI 3 vent paths does not match our information.
4 Our understanding is that the Japanese, TEPCO in 5 particular, are in the process of following and implementing 6 the emergency procedure guidelines'along the lines of whatever 7 has been developed here in the U.S. for the BRWs and those 8 guidelines do call at certain points for the use of existing 9 vent paths, if they exist.
10 We're not aware of any engineered ADPSI vent path in 11 either existing plants or the ABWR and so we will check with 12 our Japanese associates and clarify that issue in the future.
13 That's different information than we have.
14 MR. EBERSOLE: For that matter, I address this to 15 NRC, I don't think that we can guarantee an ADPSI vent path 16 either for the domestic plants. Although the APGs may say what 17 to do, in fact, I believe the physical capabilities of doing 18 that are yet to be determined.
19 MR. MICHELSON: That's another issue.
I 20 MR. EBERSOLE: I know, but it's certainly closely 21 related.
22 MR. MICHELSON: GE will give us clarification for 23 everybody's benefit, before our next meeting.
24 Thank you.
25 (Continued on page 90)
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t/ 1 MR. SCALETTI: The design, Toshiba and Hitachi 2 believe that basically, the ABWR has undergone three 3 independent design reviews, along with GD, at least for most of 4 the plants, they believe that they have provided, can provide a 5 design for the total plant, both Toshiba and Hitachi.
6 The licensing construction schedules, Stan Wilkens 7 had discussed those earlier on and they are consistent with 8 what we had found out. Originally, the plans were to provide 9 the application for the establishment permit to MITI in April.
10 MITI has a one-year review time in which they have to complete, 11 which would be I believe a document comparable to our safety 12 evaluation report to proceed with the rest of the licensing
, 13 process.
14 MR. MICHELSON: They have one year in which to do 15 their work.
16 MR. SCALETTI: They have one year. MITI has one year 17 in which to do its work, right.
18 MR. WYLIE: Do you do they have any requirements for 19 designing their plants against airplane crashes?
20 MR. SCALETTI: I don't recall. They do not, someone 21 said.
22 MR. REMICK: Dino, did you get into anything on loss 23 of offsite power? I think some of the TEPCO plants have one 24 corridor of all offsite power coming in on one set of towers,
(} 25 perhaps even outgoing power.
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() 1 MR. SCALETTI: Yes. Some of the PRAs considered loss 2 -- the PRA considered loss of offsite power, but the Japanese 3 will say they don't have station blackout because their diesels 4 don't fail.
5 MR. EBERSOLE: And do they have their own diesel 6 manufacturer and supplier as we once heard that assist in 7 reaching that conclusion?
8 MR. SCALETTI: I can't answer that. The conclusion 9 that was iterated to us was that the diesel generators do not 10 fail.
11 MR. EBERSOLE: Do not fail.
12 MR. SCALETTI: So, they are reliable and they don't 13 fail. I am sure they are maintained -- if the maintenance 14 process, we didn't see ongcing diesel maintenance while we wera 15 thers, but we certainly did see reaintenance of the turbine and 16 the rest of the reactor. And if they treat their diesels in 17 the same fashion as they do their turbines --
18 MR. EBERSOLE: I have the impression they have their 19 own diesel manufacturers.
20 MR. MICHELSON: Well, GE is going to tell us what 21 they are going to use in this case.
22 MR. RUBINSTEIN: I think your statement is valid, 23 Jesse, that as opposed to trying to adapt marine diesels which j 24 were rea31y put in place for other service, that the Japanese 25 made requirements for their diesels specifically for this
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() 1 service, which is a fast ra'pid startup, quick high load kind of 2 service and were much faster than the United States in 3 implementing things like warming circuits, pre-warming circuits 4 for oil and so forth so the diesel can be ready to go. So, I 5 think you are correct, they have arrived at reliable diesels 6 for the different process than what we have done in the U.S.
7 The second comment is that their sites have many 8 units at them and they have arranged a network of electrical
! 9 interconnections at the site so that the various emergency 10 support for one facility can come from other facilities at the 11 same site. So that the likelihood of having all the emergency 12 support for a six-reactor site be not available is incredible 13 to them.
14 MR. EBERSOLE: I can imagine that. And there are no 15 tornadoes over there either, are there?
16 MR. RUBINSTEIN: No tornadoes. There are cyclones, 17 though, of course.
18 MR. SCALETTI That is true except for at 19 Kashiwazaki, there is only one unit operating now. Eventually, 20 there will be seven.
21 MR. RUBENSTEIN: We sort of asked specifically a 22 question that said, we had some consideration of the S turbine 23 being connectable to each of the ESF diesel generated buses as 24 a AEC source to resolve station blackout concerns and to
{} 25 minimize emergency task force unavailability. Please discuss Heritage Reporting Corporation (202) 628-4888
1 l
93 g)
(_ I the Japanese rates which were respective.
2 And the answer we got is: We have EDGs, they are 3 separate and independent and they have a high availability on 4 the order of 10 to the minus 6. And, therefore, it is okay.
5 MR. EBERSOLE: Yes, I've heard that before. They 6 must be great diesels.
7 MR. MICHELSON: Does that finish up, Dino?
8 MR. SCALETTI: That concludes unless there are any 9 questions.
10 MR. MICHELSON: I see none, so, let us proceed then 11 with discussion of Chapter 4 by GE.
12 MR. CORK: I would like to introduce our next
( 13 speaker, Mr. Irv Kobsa, K-O-B-S-A. He is a principal engineer
() 14 in our mechanical equipment design group and he will be 15 addressing the reactor design, item No. 4 of the agenda and 16 item No. 5, reactor coolant system and connected systems.
17 MR. KOBSA: Consistent with the SRP, Chaptor 4
! 18 addresses the reactor assembly, the RPV and internals and I 19 will speak a little bit more about that because it is somewhat 20 covered in Chapter 5 also at the beginning, to try to give an 21 overall view of the reactor and the significant new features of 22 it.
23 Chapter 4 also contains the fuel system design and i 24 the thermal hydraulic design which for this application is the
(} 25 standard approved GE design as Dan said, anc I will not be Heritage Reporting Corporation (202) 628-4888 l
94
( 1 saying very much at all about that, except to point out the new 2 feature we have in the control rod, itself. And we will get 3 into reactor materials and the reactor control system. In 4 particular, the details of the new control rod drive.
5 (Slides shown.)
6 In coming up with this reactor vessel and internals, 7 our design objectives were as outlined here to define a reactor 8 system which would fully contribute to all the major objectives 9 of the ABWR program, simplification of design, trying to 10 provide additional margins in that design with reduced 11 maintenance cost and reduced maintenance that would result in 12 lower exposure to the operators, and capacity factor and
. . 13 availability of improvements and enhanced reactor safety. It
- 14 would also address all known BWR material concerns and use 15 established technology from the different reactors in the world 16 or the new feature would be tested prior to its application and 17 use.
18 I had thought that we would have two projectors, one 19 that I could show you these design features in an outline form 20 and also point them out on the viewgraph of the reactor, 21 itself. So, what I will try and do is go through --
22 MR. MICHELSON: Would you like to have two 23 projectors? We can arrange for another one.
24 MR. KOBSA: No. That's okay.. You have the words in 25 the front of you and I will use that as a guide in illustrating
(]}
Heritage Reporting Corporation (202) 628-4888
95 x' 1 'and showing you this on the reactor.
2 MR. MICHELSON: We have got two projectors.
3 MR. KOBSA: This will do quite well.
4 MR. MICHELSON: Will you need it this afternoon? If 5 you would like it, we will arrange it.
6 MR. CORK: It would help our presentation if we were 7 to have it.
8 MR. MICHELSON: We will make sure we have got two as 9 soon as it's available.
10 MR. KOBSA: This is the reactor vessel arrangement.
11 The vessel is very similar to the BWR 6 design in general. The 1.? reactor internals really fairly well dictating the design 13 arrangement. The vessel diameter is larger than the largest O 14 BWR 6 that we have had. It's 27 inches larger, primarily, to 15 allow space for the reactor internal pump removal.
16 The reactor pressure vessel does have a reduced 17 diameter closure rather than other reactor vessels that we have 18 had having a straight full diameter opening. This reactor h.as 19 a reverse flange and that is possible because we don't require 20 that full area for steam separation equipment. And, therefore, 21 have reduced the diameter of the opening and the seal, thereby, 22 reducing the number of bolts and the size of 'the head that has 23 to be removed. This results in a fairly significant reduction 24 in the cost of the head and also in the time required for the 25 bolting and unbolting.
(])
Heritage Reporting Corporation (202) 628-4888
96 en k-) 1 The reactor internal pump penetrations are in the 2 knuckle region of the bottom head and that knuckle is formed in 3 one large forged ring. And the penetrations are machined into 4 that forging. There is no welding like a convention nozzle 5 involved. It is machined from the forging.
6 The steam flow restrictor is integral in the nozzle 7 of the vessel. Now, this is a new feature for us. It is a 8 feature that has been used in European reactors and is built 9 into this vessel.
10 The RPV head vent and spray nozzle in the top head, 11 we have always had a vent and a head spray, but this spray is a
[ 12 little bit different in that 1c sprays and washes the vessel 13 head and flanges, themselves, rather than spraying the steam 14 space and just condensing the steam as we have done in the 15 past.
16 MR. REMICK: What is the advantage of putting the 17 flow restrictor in the nozzle?
i 18 MR. KOBSA: By having the flow restrictor in the 19 nozzle, this reduces the pressure loading across the reactor 20 internals in the event of the postulated steamline break. It 21 also reduces the pressure break buildup in the containment in 22 the case of the postulated break.
23 MR. REMICK: I guess I don't understand how putting 24 in the nozzle is any better than putting it --
25 MR. KOBSA: In the line?
(])
l Heritage Reporting Corporation l (202) 628-4888 l
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N- 1 MR. REMICK: In the line.
2 MR. KOBSA: Well, you could have a break between the 3 restrictor and the vessel. That was the worst break.
4 MR. EBERSOLE: On the matter of the core flood -- at 5 one time you had to claim spray distribution, but you 6 eventually discarded that and said you have put it in the upper 7 head to the point where liquid entry was accomplished by a 8 simple overflow into the several channels. Are you defending 9 any more on spray distribution per se?
10 MR. KOBSA: No, we are not. Craig will talk about 11 that in more detail this afternoon. But, no, we are not.
12 As far as the reactor internals are concerned, then, i
13 the BWR 6 type steam dryer and steam separator are reapplied 14 here. These have been in performance for us in the BWR 6's and 15 as I mentioned we have enough space so they can easily handle 16 the flow of the ABWR in the space provided.
17 We have the reference design for the control blade at 18 this time is a boron carbide blade. It does not have a 19 velocity limiter. I will show you a picture of that in a 20 moment and that makes it shorter and less weighty for l 21 insertion, and easier to insert.
l
! 22 MR. REMICK: How did you get away from the flow 23 limiter or the velocity limiter?
! 24 MR. KOBSA: It is because of the control rod drive 1
25 mechanism. And I will talk about that in detail at the drive.
({')
Heritage deporting Corporation (202) 628-4888 l
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'9 8 -
1 It prevents separation from occurring.
2- There-is a separate low pressure flood sparger in=the 3 vessel and these are arranged just below the feedwater 4 spargers. There is two of them and there are three total 5
' systems, one of them comes through the feedwater,'two of them 6 through two-separate spargers inside the vessel. And then 7 those are low pressure flood spargers. And there is a high 8 pressure flood'sparger, two that are similar to these and are 9 installed in annular sections in the top guide below the shroud 10 head so that they are permanently installed or they don't come 11 out with.the shroud head. They stay there during refueling.
12 MR. EBERSOLE: What in the importance of the sparging 13 process versus just simple introduction of water into the top 14 head?
15 MR. KOBSA: Well, as opposed to jetzing a big stream 16 of water, this distributes it more uniformly and then doesn't 17 concentrate .s cold and high jet --
18 MR. EBERSOLE: But it has been often thought that 19 thermal shock will break those pipes and you might lose them 20 anyway, so, you could depend on the jet anyhow. Is this 21 correct?
22 MR. KOBSA: No, no. Far from it. The spargers are 23 designed and analyzed under ASMB Section 3 rules and predicted 24 to be well within the --
i O :s MR. EBERSOLE: Even when you hit them empty with the Heritage Reporting Corporation (202) 628-4888
f- 99 I')
' 1 cold slugs of water?
l 2 MR. KOBSA: Yes.
l 3 MR. EBERSOLE: Okay.
4 MR. MICHELSON: The ASME doesn't address water,
[ 5 slugs.
! ~
[ 6 MR. EBERSOLE: That's true. But why not a single 7 nozzle? Why this whole elaborate array if you don't need a l
L f 8 spray distribution.
9 MR. KOBSA: Well, it is really not just such a whole 10 elaborate array.
l 11 MR. EBERSOLE: It is a complete ring header, I l
12 assume.
l 13 MR. KOBSA: No, it is not. These are segments. The 14 feedwater sparger is in six equal length segments around the '
15 vessel. In other words, about 60 degrees.
16 MR. EBERSOLE: I am thinking of inside the shroud.
17 MR. KOBSA: Well, inside the shroud, the same kind of 18 thing. And the flooding spargers are really virtually 19 identical to the feedwater spargers except that the elbow 20 outlets at the top are smaller in diameter because the system 21 flow is smaller.
22 And the sparger on the inside is essentially the same 23 as a flooding sparger. It's got I think it's 18 elbow outlets 24 on it. It is not like the core sprays that we have had in the
() 25 past that go completely around. It is a segment of less than Heritage Reporting Corporation (2'02) 628-4888
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(i !
1 90 degrees each. So, there is two of these in two segments of ;
2 ring sparger that are less than 90 degrees around.
3 MR. EBERSOLE: If you need distribution like that, 4 how do you check to see that you still have it?
5 MR. KOBSA: Distribution is not a requirement in 6 these systems. It is just a matter of getting the water into 7 the reactor.
8 MR. EBERSOLE: Well, that is what I wanted to hear 9 you say, but I didn't think I heard it before.
10 MR. KOBSA: I'm sorry I didn't make that clear. No.
11 The distribution is not at all required. It is because the 12 core never uncovers. The core remains covered.
13 We have a in core guide tubes and a network of bars C 14 that support them and prevent their vibration in the 15 recirculation flow stream that is in the reactor. We have had 16 a similar structure to that in past reactors. The additional 17 feature here is that those supporting bars are tied off to the l
18 shroud and shroud support to give them a greater fixity. This 19 we think necessary because of the higher LOCA velocity flows in 20 the bottom head coming from these pumps.
21 MR. MICHELSON: Is there some particular reason why, 22 since you don't have recirculation lines anymore to break that 23 you have to put your sprays inside of the shroud area? The 10 24 pumps at the bottom are all bypasses --
() 25 MR. KOBSA: No. As a matter of fact, we had i
i Heritage Reporting Corporation (202) 628-4888
101 I) s- 1 considered not putting it in there and this, I think, is mainly 2 for a diversity of --
3 MR. MICHELSON: It is a little better inside isn't 4 it? If you really believe you never have any way of draining 5 the vessel, then outside would be perfectly all right.
6 Wouldn't it.
7 MR. KOBSA: Correct.
8 MR. MICHELSON: Just another flooder.
9 MR. SAWYER: It's a close call. We decided that we 10 would like to have at least some of the water added on upon the 11 inside even though distribution isn't necessary.
12 MR. MICHELSON: Just in case you lose one of those 13 pumps catastrophically, that's one time it would be nice to O. 14 have it this way.
15 MR. KOBSA: Well, one of the things that we have done 16 to mitigate or prevent that from happening is that the pump the 17 way it is configured, the shaft of the pump backseats against 18 the housing, and, also, we have restrainte on the pump casing 19 to prevent the separation of the pump if it is assumed to fail.
20 MR. EBERSOLE: Have those pumps brought about some 21 new nozzle problems to enhance those already presented by the 22 control rod drives? The structure of the bottom of the vessel
, 23 gets to be a rather complex -- you said that was a single 24 forged and they machined the holes.
25 MR. KOBSA:
(]) The knuckle section. I am giving Chapter He.vitage Reporting Corporation (202) 628-4888
102 0 1 5 presentations after lunch. But there is only a single weld 2 in the bottom head and that is a circumferential weld that goes 3 between where the pumps penetrate the head and where the drives 4 penetrate the head. There is no other weld there. And, so, 5 this is made as a ring forging, rather similar to the ring 6 forgings that are like the vessel flanges and that contour.
7 MR. EBERSOLE: Who makes this vessel? The Japanese 8 only?
9 MR. KOBSA: Yes.
10 MR. EBERSOLE: So, that is a Japanese product.
11 MR. KOBSA: They are the only ones that have the 12 capability for making that sized forging and that sized vessel.
13 MR. EBERSOLE: One place. One source.
14 MR. KOBSA: Well, no. There is at least the two 15 sources, both Hitachi and Toshiba, their companies have 16 capabilities to fabricate this vessel and these forgings can be 17 procured from Japan only. Japan steel.
18 MR. EBERSOLEr. So, that is a Japanese vessel.
19 MR. KOBSA: World class vessel.
20 MR. WYLIE: Are you going to say something about how 21 many individual welds you have got around the vessel?
22 MR. KOBSA: I can. That might be best taken up in 23 this Chapter 5 presentation.
24 There is, of course, the different control rod drive.
() 25 It doesn't show particularly here, but let me point out that Heritage Reporting Corporation l (202) 628-4888 I
s 103 1 there is the electro-mechanical control rod drive and what you 2 see extending below where the control rod drive normally ends 3 is the electric motor and the syncros that measure its 4 location.
5 The last feature noted here was that the feedwater 6 nozzle is in the ABWR a welded design. The thermal cleeve is 7 welded into the safe end of the nozzle. It does have e second 8 thermal sleeve outside of the primary thermal sleeve for added 9 protection to the feedwater nozzle from thermal fatigue.
10 MR. EBERSOLE: I see you have a control rod drive 11 restraint beam. That is the same thing you have on the present 12 plants; isn't it?
13 MR. KOBSA: Yes. That is the same. That's a lateral
(~)N
\m 14 restraint.
15 MR. EBERSOLE: Not a vertical thrust?
16 MR. KOBSA- No.
17 MR. EBERSOLE: In other words, have you lost any 18 control rod ejection problem?
19 MR. KQESA: Well, no. We have loct all that 20 undervessel steel, as it is sometimes called, and cleaned that 21 up by virtue of -- and I will show you the details of that in a 22 moment with the control rod drive. The drive, itself, couples 23 into the control rod guide tube which in turn is rested on the 24 CRD housing. So, it is held in redundantly to the CRD housing
() 25 by virtue of its coupling to the reactor internals. There is a Heritage Reporting Corporation (202) 628-4888
104
. S m) 1 bayonet coupling in there so that it is restrained from blowout 2 but not by the separate undervessel steel as before.
3 MR. EBERSOLE: You argue now you have eliminated the 4 blowout problem?
5 MR. KOBSA: No. We have eliminated the ejection from 6 the -- the rod drop problem.
7 MR. EBERSOLE: No, I am talking about the control rod 8 housing blowout problem.
9 MR. KOBSA: No. We have designed features to protect 10 against that.
11 MR. EBERSOLE: Well, that used to be a grid.
12 MR. KOBSA: It did, yes. Well, yes. There's hangers
{} 13 from a structure in there and that's right.
used to be. It is not there now.
That's what it That structure is removed 14 15 making maintenance and removal and reple. cement of the drives 16 easier and less time consuming.
17 MR. SAWYER: The problem is selved a different way.
18 MR. EDERSOLE: Okay. We'll get into that.
19 MR. KOBSA: It is not that we have said that that 20 can't happen.
21 As I mentioned in the overview, the fuel system is l
22 approved fuel design separately addressed and approved by NRC 23 and I suppose that by the time an ABWR would be produced would
- 24 be different than that, but basically the SELAD is 8 by 8 fuel.
() 25 And the one thing that I would point out is this is the design l Heritage Reporting Corporation l (202) 628-4888 '
l
l 105 1 of the control rod. The control rod reference design they were 2 working with is still with multiple vertical absorber rods 3 encassd in a sheath which is the control rod we have had right 4 along, but the velocity limiter is no longer at the bottom.
5 And, as a result of not having a velocity limiter at the 6 bottom, that makes.the control rod about a foot shorter which 7 contributed to the overall shortening of the reactor vessel.
8 MR. EBERSOLE: You had to do something to do that.
9 Did you make some sort of a detection system?
10 MR. KOBSA: Yes, exactly. And so we have in the 11 drive, and I will show you the way that that works with the 12 drive mechanism of detecting whether or not the control rod is 13 following it.
(]}
14 Tho reactor materials as addressed here in Chapter 4 15 of the reactor internals materials are almost all austenitic 16 stainless steel. They are exclusively type 304L or type 316L 17 per ASME specifications if they are for the core supporting 18 structure or ASTM specifications if they are for the internal 19 structures which are not coated components, not core 20 structures.
21 The exceptions to that are that the shroud support 22 ntructure, that is the structure which makes the transition 23 between the shroud and the vessel is nickel chrome iron 24 inconel 600, as are the shroud head bolts which are used to
() 25 secure the shroud head on the reactor.
Heritage Reporting Corporation (202) 628-4888
106 V(s 1 MR. EBERSOLE: That inconel 600 material has been 2 expe lencing some stress corrosion cracking in BWRs I believe.
3 Am I correct?
4 MR. KOBSA: Yes.
5 MR. EBERSOLE: Is that correct?
6 MR. KOBSA: There has been casas of inconel 600 7 stress corrosion cracking and we have a particular criteria 8 which we have adopted to prevent stress corrosion cracking of 9 both staialess steel and the inconel 600. We have devised some 10 special stress rules that we apply to that.
11 MR. MICHELSON: Do you think -- of course, you have 12 got to be careful with heat treatments and a number of other
(; } 13 things, too, but you think now you have licked the stress 14 corrosion problem as it relates to reactor vessel attachments?
15 MR. KOBSA: Yes, sir.
16 MR. MICHELSON: And you are still using inconel 600?
17 MR. KOBSA: Yos. And there has not been any sign of 18 a problem of stress corrosion cracking in these shroud supports 19 which we have been using --
20 MR. MICHELSON: I thought the attachments for those 21 core support members to the vessel are where we are seeing the 22 cracks, but that's another subcommittee which I attended just 23 the other day. And I wondered how the ABWR was going to 24 approach that.
m
(-) 25 I guess this is where we are going to get into the Heritage Reporting Corporation (202) 628-4888
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1 materials.
2 MR. WILKENS: Let me suggest that at some point, it 3 would probably be appropriate for us to bring our materials 4 here. ,
5 MR. MICHELSON: Yes. The next subcommittee meeting 6 when we cover questions, we will tell you the kinds of things 7 we are interested in on this subject and then we will go back 8 and pick up the material questions.
9 MR. WILKENS: I can tell you that everything in the 10 ABWR design has been done to everything we know about BWR 11 materials performance, but we really need to get our experts 12 here in that area and walk you through the details of that.
13 MR. MICHELSON: The cracks have been appearing in the
(])
14 vessel metal, but coming from cracks propagating by stress 15 corrositn in the inconel 600 attachments in some cases. And I 15 just wendered on that whole issue which I guess is a fairly new 17- issue, that we certainly would like to know why the ASWR 18 doesn't have such problems. And we will get into that in the 19 next session.
20 (Continued on next page.)
21 22 23 24 25 Heritage Reporting Corporation (202) 628-4888
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/~)x N_
1 MR. KOBSA: In addition to inconel 600 we are using 2 in the hardware the small bolts that are used to mount the core 3 plate and the top guide and the bolts and pins there are not 4 welded, they are either 304, 316 or actually primarily XM-19 5 which is the nitronic 50 the higher strength, the nitrogen 6 addition stainless.
7 And the core support structures are of course 8 constructed in accordance with the ASME code and are fabricated 9 in accordance with Section 3 and Section 9 of the code.
10 The fabrication and processing of the austenitic 11 stainless steel does comply with the applicable regulatory 12 guides.
13 And as indicated here, the base metal of the reactor
(])
14 internals is the L grade and for us that means typically a .020 15 percent carbon. And it is annealed and verified that it has 16 been solution annealed by testing for sensitization.
17 Cold forming is controlled and limited by radium of 18 bends and by a hardness test applied to the final, finally 19 formed products, with the exceptien of the dryer veins.
20 The dryer veins have always been the thin sheet cold 21 formed stainless steel.
22 And of course, this does prevent the yield strength 23 of the material from even approaching the 90 KF. ilmit 24 mentioned in the standard review plan.
O 25 The weld metal is selected and controlled to control Keritage Reporting Corporation (202) 628-4888
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1 the delta ferrite content as indicated there with a ferrite 2 number of 8 on the average and 5 minimum to prevent hot 3 cracking and also in'tergranular stress corrosion.
4 And the use of low carbon material is the main thing 5 that prevents sensitization and a problem, but also weld heat 6 input control contributes to that and has been effective in 7 preventing these problems.
8 MR. MICHELSON: Were you intending to tell us where 9 your welds are in the higher radiation zo'nes in the vessel?
10 The way to avoid some of our problems that have no 11 welds --
12 MR. KOBSA: As a matter of fact, in the vessel
() 13 itself, I mentioned that in connection with Chapter 5, you'll 14 see that we do have figures there for the effects of radiation 15 on weld metal but we -- and we had, as this report was 16 prepared, our design did not preclu.de initially welds that 17 wauld be in the belt line, but are depigned as of now, would 18 follow the EPRI guidelines an the ALWR requirements. And we 19 will not have welds in the belt line at all.
20 The belt line will be formed by a single ring forging 21 so that there will be no weld metal in the belt line.
I 22 MR. MICHELSON: How high a forging is the --
23 MR. KOBSA: It's not very much more than the core 24 length. It's only a few inches above and below the bottom of 25 the core.
Heritage Reporting Corporation (202) 628-4888
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1 But at that location, the fluents at the surface is i 2 less than 10 to the 18th.
3 MR. EBERSOLE: The bottom of the vessels is always an 4 interesting geometric and stress problem, and attachment 5 problem.
6 We'd like, I'm sure, to see a rather detailed picture 7 of how the control rod housings are made leak type without 8 possibilities of cracks and how they won't spit o':t. And now 9 of course it's compounded by what I understand are forgings for 10 the pumps.
11 MR. KOBSA: I think I can point that out on the 12 figures for both the drive and the pump for you, best.
13 MR. EBERSOLE: I think one thing is it is hard to
(])
14 fix any problems that might come down there. So you want to do 15 it right the first time.
16 MR. KOBSA: Okay. And of course, cleanliness I 17 guess, what's the saying, being next to Godliness? Special
. 18 care is exercised to avoid the contamination during the 19 fabrication and shipping and installation of the racked 20 materials. Any material that could contaminate and be 21 contributory towards stress corrosion cracking is precluded 22 from contacting or if it is inadvertently in contact is 23 immediately cleaned from the material.
24 MR. MICHELSON: While you are at an interrupting d 25 point, let me go back for a moment and ask you, the fuel Heritage Reporting Corporation (202) 628-4888
(')
v 111 1 application is NDE 24-011. Has that been submitted to the 2 staff yet, do you know?
3 It's your General Electric Standard Application for 4 reactor fuel.
5 MR. SAWYER: Joe, you know that a G-Star uptlate has 6 been submitted that includes the ABWR7 7 MR. CORK: I think the answer is yes it has and-I 8 think there is some additional information that we're going to 9 provide at a later date. But yes, the G-Star amendment for l0 ABWRM.
11 MR, MICHELSON: Because the review of Chapter 4 is 12 heavily referenced to that.
13 MR. SAWYER: Referenced to G-Star, that's correct.
(])
14 MR. MICHELSON: And the staff would certainly be 15 reviewing it at the same time to the extent that it hasn't 16 a.'_ ready reviewed it.
17 That leads up to another question, Dino. How do we 18 know which one of these documents has already been reviewed?
19 'Like the fuel document must have been reviewed for G-Star.
20 MR. SCALETTI: For G-Star?
21 MR. MICHELSON: Well, this 24011 I thought was not a 22 brand new document, or is it?
23 MR. SAWYER: No, it's not. It's a document that 24 includes all of the active fuci designs that are being applied 25 at any BWR.
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1 What we did was we added the ABWR-as an option ~.
2 MR. MICHELSON: So you probably reviewed it already.
3 MR. SCALETTI: I don't think we reviewed it. .
4 MR. MICHELSON: Oh, you haven't.
5 MR. SCALETTI: I do recall something coming in about 6 three months ago.
7 MR. MICHELSON: That's another whole review problem.
8 MR. SCALETTI: I'll check on that.
9 MR. MICHELSON: It's a fairly extensive document.
10 Okay. Thank you. Go ahead.
11 MR. KOBSA: The final section of Chapter 4 deals with 12 the reactivity control systems and the design basis for these
() 13 systems, or from the safety standpoint, to provide for rapid 14 insertion of the control rods, the scram function, and these c 15 control rod drives are comparable. They have essentially the 16 samo scram response time as our fast scram locking piston drive 17 used on BWR-6.
18 The control rod drive does of course position and 19 individually support and individually position the control rod 20 itself. The drive does prevent rod drop through separation 21 detection which I'll show you how that's accomplishad. And 22 that's the reason then that the control rod velocity limiter 23' can be eliminated. And it does prevent or limit the rate of 24 rod ejection. And that's to your question on the ejection. In O 25 the event of pressure boundary failure through redundant means, Heritage Reporting Corporation (202) 628-4888
113 1 a break, which is actually now electro-mechanical break, the 2 report actually' described a centrifugal break but we've changed 3 to an electro-mechanical break which is more testable than the 4 centrifugal break that we had planned initially, and a check 5 valve in the supply line prevent any pressure boundary failure 6 in the drive housing or the drive itself that could cause this 7 ejection and rapid blowout of the drive and control rod.
8 From the power generation side, the increment of 9 control is smaller and of course the electric motor positioning 10 of it makes it more adaptable to automatic controls.
11 MR. EEERSOLE: Has this now led to the ability to 12 gain withdrawal?
() 13 MR. KOBSA: Yes, it does.
14 MR. EBERSOLE: And stop this daisy chain process of 15 restart?
16 MR. KOBSA: Yes, it does.
17 MR. EBERSOLE: It enhances restart by number of 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />?
19 MR. KOBSA: Yes, that's one of the key advantages of 20 the drive system.
21 MR. EBERSOLE: Do you still use pump trip as a pre-22 activity control for Atlas concepts?
23 MR. SAWYER: Your conclusions in all that are correct 24 in the starting. We use up to 26 rods in the tying and O 25 covertical and then we switch to 8 and 4 rods. So that does, F s Heritage Reporting Corporation (202) 628-4888
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114 1 we are trying to be able to do a hot restart in about four 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
3 MR. EBERSOLE: Are these new pumps high frequency, 4 high RPM pumps, with higher frequencies?
5 MR. SAWYER: There are 1,500 RPM pumps, high speed.
6 MR. EBERSOLE: ksee. So it's the old 60-cycle 7 system.
8 MR. KOBSA: That covers through power aupplies.
9 These are the features of the control rod drive and 10 I'll go over those and describe them to you with the aid of the 11 picture schematic that you have as the next chart there in your 12 package.
() 13 In concept and in execution the control rod drive is 14 really a very simple design. It may not quite look it from 15 that even schematic there.
16 But basically it's like two pipes, one within the 17 other, that the upper pipe being the hollow piston connected to 18 the control rod, the lower pipe, the fixed pipe, being the 19 outer tube of the drive.
20 And within that, there is a ball screw and nut that 21 is rotated by a motor below, a stepping motor, that positions 22 the nut.
23 The nut of course advances on the screw because it is 24 prevented from turning in a guide tube inside of the drive.
i
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1 and that raises and lowers the inner pipe. And the inner pipe 2 is the hollow piston of the control rod, it is the part 3 connected to the control rod and is the pos'itioner.
4 MR. EBERSOLE: Then you no longer have to have flow 5 throughput to keep the nozzle cool, do you? Or do you?
6 MR. KOBSA: Well, it wasn't really to keep the nozzle 7 cool but to keep the drive cool and the seals from getting 8 overheated. That's right. We don't need it for that reason 9 anymore, because there is no longer a seal, graphite tar seals 10 in this drive like there has been in the past.
11 MR. EBERSOLE: Where is the critical seal weld, the
, 12 critical fixed seal weld that keeps the whole thing from
() 13 leaking? The one which one it cracks is chaos.
14 MR. KOBSAs That would be right here, with the CRD 15 housing drawings, the stub nozzle and the vessel.
16 MR. EBERSOLE: That's the weld that better not crack 17 and cause leaks because of just the annoyance if nothing else.
1 1
18 MR. KOBSA: That's right. That's a very difficult 19 weld to, would be a difficult weld to repair.
20 MR. EBERSOLE: Right. Is there anything other than 21 just a plain seal weld there?
22 MR. KOBSA: Oh, yes. That's a strength weld.
23 MR. EBERSOLE: A strength weld to boot.
24 MR. KOB3A It's an ASME code recognized weld. It's O 25 a combination groove and fillet weld that is -- it has greater Heritage Reporting Corporation (202) 628-4888
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1 cross section than the tube itself.
2 MR. EBERSOLE: So the strength weld is also a seal 3 weld?
4 MR. KOBSA: Yes. I don't think of it particularly as 5 a seal weld. There isn't a separate seal weld. It's a 6 combination groove and fillet weld specially recognized in the 7 code for this kind of application.
8 MR. EBERSOLE: It's just space seals which are 9 removable.
10 MR. KOBSA: Pardon?
11 MR. EBERSOLE: Through the rest of the leakage 12 problem, are face seale which you can pull out and fix.
() 13 MR. KOBSA: Right.
14 MR. MICHELSON: The in-vessel housing support, is it 15 actually welded also by that lower full penetration weld?
16 MR. KOBSA: That detail design has not been finalized 17 but what this illustrates is a ring that is welded to the upper 18 part of the CRD housing and that's there strictly for -- it is 19 not normally loaded at all, but it is strictly for the 20 eventuality that if this weld failed by a path which we have 21 never been able to make it fall and test or analysis doesn't 22 tell us it would fail that way, the failure path of analysis 23 would predict and tests have demonstrated is across the CRD 24 housing itself.
'- ' But if it failed along the housing, then the whole 25 Heritage Reporting Corporation (202) 628-4888
'(O 117
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1 CRD housing could be, with the drive and blade attached, could 2 be ejected.
3 So this'is there to prevent that occurrence from t- 4 possibly happening.
5 MR. MICHELSON: But the housing is not actually 6 welded to the nozzle, is it?
7 MR. KOBSA: Oh, yes. The housing is that part of the 8 nozzle right here. That's a pressure boundary strength weld.
9 MR. MICHELSON: Okay. That's a full penetration weld 10 that involves the housing as well?
11 MR. KOBSA: it's not a full penetration weld. It's a 12 weld that is a groove and fillet wet.d.
m k_) 13 MR. MICHELSON: I don't know how you'd make that 14 weld, if it's also welding the housing at the same time, unless 15 you did make it full penetration.
16 MR. KOBSA: The housing is a forged housing --
17 MR. MICHELSON: -- housing support, I'm sorry.
18 MR. KOBSA: The housing support?
19 MR. MICHELSON: Yes.
20 MR. KOBSA: The housing support in this concept that 21 is shown is a separate loose ring that is put on the housing 22 after the housing is installed and welded in and it is welded.
23 MR. MICHELSON' That was the question. Was it welded 24 to the nozzle.
25 MR. KOBSA: No, it's not welded to the nozzle. Only Heritage Reporting Corporation (202) 628-4888
) 118 .
1 to the housing.
2 MR..EBERSOLE: What are the comparative stresses at 3 the bottom of your vessel compared the stresses in the outor 4 perimeter of the vessel? Are they higher here?
5 MR. KOBSA: The mean stress is the same.
6 MR. EBERSOLE: Is it?
7 MR. KOBSA: Yes. The head is 11 inches thick. It's 8 that thickness to control that very parameter. So that the 9 mean stress between the weld, between the openinos for the 10 drive and the rest of the vessel is the same mean stress.
11 MR. EBERSOLE: How far is it in the webbed section 12 there to the next rod?
r's
(/ MR. KOBSA: The center to center distance is 12.2 13 14 inches. Two tenths of an inch more than we've used in the 15 past.
16 MR. EBERSOLE: And that leaves perimeter to perimeter 17 of the rods -- how far is the perimeter of the rod housings 18 apart?
19 MR. KOBSA: The perimeter of the rod housings?
20 MR. EBERSOLE: Yes.
21 MR. KOBSA: The distance between holes is 6.2 inches.
22 MR. EBERSOLE: Okay. Thank you.
23 MR. KOBSA: Just a little bit more than half the 24 distance. That's essentially the same as it's been. It's been 25 six inches, six inch holes on 12-inch centers in the past. The Heritage Reporting Corporation (202) 628-4888
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() 119 1 extra two tenths of an inch was for fuel performance core 2 considerations, more optimum.
3 MR. EBERSOLE: Six inches apart but it's ten inches-4 thick. >
5 MR. KOBSA: Eleven.
G MR. EBERSOLE: Eleven inches think.
7 MR. KOBSA: Minimum, yes.
8 The control rod, the main features of it are these 9 that I described that the outer housing contains the pressure 10 coming from the hydraulic system the accumulator, for scram, 11 and the only seal that there is a labyrinth seal, on the -- a 12 stationery seal attached to the upper part of the outer tube,
() 13 so that this hollow piston moves through that labyrinth seal 14 and a whole drive then under that labyrinth seal is pressurized 15 from the hydrauli.c system and the effective area that is tie 16 hollow piston area passes through this labyrinth seal.. Thit's 17 the effective area that causes the pressure of force from the 18 accumulator to insert the rod capidly on scram.
19 MR. EBERSOLE: There is a continuous leakage past 20 that seal, isn't there?
21 MR. KOBSA: Yes, there is. There is a very small --
22 MR. MICHELSON: How much is that?
23 MR. KOBSA: Three tenths of a gallon a minute. ;
24 MR. MICHELSON: And so there is a continuous CRD pump 25 requirement for flow equal to that time the number of rods?
Heritage Reporting Corporation (202) 628-4888
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1 MR. KOBSA: Right.
2 MR. MICHELSON: Now that is also a cooling flow 3 essentially, or could be..
4 MR. KOBSA: Yes, it is.
5 MR. MICHELSON: And is it required to be there for a 6 cooling flow?
7 MR. KOBSA: No, it is not, n
8 MR. MICHELSON: Now is it required though to protect 9 the equipment not necessary to protect the pressure boundary?
10 MR. KOBSA: The continuous purge flow there is to 11 prevent contamination of the drive, to prevent any settling of 12 radioactive material from the reactor down into the drive pm
%_) 13 mechanism.
14 MR. EBERSOLE: If that core stopped up for any 15 reason, what's the consequence?
16 MR. KOBSA: It reduces the stresses in the housing.
17 It makes it more isothermal and --
18 MR. EBERSOLE: But it doesn't impede the rod action?
19 MR. KOBSA: No.
20 MR. EBERSOLE: Does all the water come from the same 21 source in which you used to think about might be polluted with 22 the gelatin, you know?
23 MR. KOBSA: From the condensate storage?
24 MR. EBERSOLE: Yes, burst filters and so forth?
O 25 MR. KOBSA: Well, the primary source of water for the Heritage Reporting Corporation (202) 628-4888 l
- -__ _ _ - _ - - _ _ - - - - _ _ )
121 1 drive system is the condensate storage tank. It's 2 demineralized water.
3 MR. EBERSOLE: Do you eliminate the concept of 4 simultaneous plugging of this with any substances by filtration 5 systems?
,6 MR. KOBSA: There is a filter in the line to the 7 drive system, yes. But that's not an essential system. That 8 whole drive supply system is a non-essential system. The 9 accumulators are precharged and they of course perform the 10 scram function.
11 MR. EBERSOLE: I understand that but I'm thinking 12 about plugging up the mechanics of the system with some
() 13 contaminant which used to be the gelatin --
14 MR. KOBSA: The resins.
15 MR. EBERSOLE: The resi-3. Yes.
16 MR. KOBSA: Demineralized resins. That's happened.
17 MR. MICHELSON: That will proceed even though you are 18 scramming. It builds up in time.
19 MR. EBERSOLE: You don't have any duality of flow 20 from separate sources or anything like that? How do you know 21 that that water is always clean that you keep pumping through 22 these things? And you pump all from the same source, don't 23 you?
24 MR. KObSA: I'm not sure what monitors are on O 25 condensate storage.
j Heritage Reporting Corporation (202) 628-4888
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1 MR. EBERSOLE: It's got to be mighty clean when it 2 goes to the rods. It wouldn't matter a lot to the fuel because 3 it would be a gradual situation, i
4 MR. MICHELSON: This opens a whole set of questions.
5 MR. EBERSOLE: I know it.
6 MR. MICHELSON: And I wanted to ask General Electric, 7 and this is as good a time as any. Where can I read about the 8 failure modes and effects of these control rod drives, 9 including both the air side of this control system and the 10 water side? I found no reference to any, I found no discussion 11 of it. I assume the staff is certainly interested in how these 12 things get screwed up from air contamination or from water
() 13 contamination. And so where is General Electric going to tell 14 us about what thinking you've given to these kinds of questions 15 Jesse just asked, and there are a number of others as well?
16 Certainly you must have gone through some kind of a 17 failure mode on the supply side to see that the water supply 18 wouldn't screw it up or the air supply wouldn't screw it up, or 19 the electrical to the control rod modules.
20 MR. CORK: Do you know of any?
21 MR. KOBSA: I don't know --
22 MR. MICHELSON: I would certainly expect before we're 23 done to read a document that describes why General Electric 24 thinks this design is immune from these kinds of problems.
O 25 MR. KOBSA: This drive, not like the locking piston Heritage Ref :! ng Corporation (202; -n-4888 i
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123 1 drive, which is positioned both for shim and scram 2 hydraulically, is scra.nmed, is hydraulically scrammed only and 3 that makes it much less vulnerable to things like resins or any 4 other particular matter, or anything for that matter --
5 MR. MICHELSON: If you contaminate the bushing of the 6 seal area badly enough it certainly might retard the motion of 7 the drives.
8 MR. KOBSA: It might. From our testing, and our 9 concerns are more the other way around.
10 MR. MICHELSON: If I put too high a voltage on those 11 scram valves and I boil out the varnish in the valves, you 12 might not scram. If I put enough water or contaminants into
(--) 13 the air system, you might have a problem. These are things ,
14 that we want answers as to why these won't be a problem.
15 MR. CORK: Yes. The incidences that you're talking 16 about have occurred in the field and associated with those fuel 17 events or fixes, and in the case that you mentioned, of 18 contaminant, that was a urethane disc problem which has been 19 replaced by a material that no longer has that problem. And 20 that is an ABWR design.
21 MR. MICHELSON: See, what we'd like to do though is 22 read somewhere where GE has gone through these potential 23 sources and then the valves were just one possible source of
{) 24 25 this problem. You also had some boilout of varnish on the solenoid side of these valves at Pilgrim I believe and a couple l
l Heritage Reporting Corporation (202) 628-4888 ,
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() 124 1 other places. But I would think that there would be a 2 document, a failure modes and effect analysis done on this-kind 3 of a drive system, since it's so critical. And I didn't see 4 any reference to any and I was going to ask, do you think there -
5 will be such a document or do you just supply this by people's 6 knowledge and remembrance, or whatever? There's got to be a 7 documentation with tables on something as critical as control 8 rod drive, would think.
9 MR. RUBENSTEIN: Could you quickly run by the 10 hydraulib scram, how the piston rises? What's the path of the 11 scram hydraulic fluid to lift the piston? Is that through the 12 ball nut?
13 MR. KOBSA: Yes. Well, not through the ball nut but 14 around it.
15 The water enters from the accumulator in this line.
16 There is a check valve here. If you are familiar with the pass 17 drives, this check valve served a couple functions. But it 18 serves only a function now to block this line in the event of a 19 line failure.
20 If the passages from that line go into the, inside 21 this outer tube and under the hollow piston, so that the --
22 MR. RUBENSTEIN: 'Are there any special provisions for 23 the liquid to get under that piston? How well does it seal 24 with the ball nut? That's what's confusing me.
25 MR. KOBSA: No, it doesn't seal with the ball nut at Heritage Reporting Corporation (202) 628-4888
() , 125 1 all.
2 MR. RUBENSTEIN: Okay.
3 MR. KOBSA: There's no seal there, this rests on the 4 ball nut and it provides a platform to move this in and out but 5 the filling of the outer tube pressurizes the area of the 6 hollow piston which is a small area that the pressurs is above 7 and below and the net area of the hollow piston that passes 8 through the labyrinth sees the full effect of the pressure -
9 that's contained in the full drive. So this whole cavity which 10 is the outer tube up to the labyrinth seal is pressurized. And 11 the area then which is not pressurized is the top of the hollow 12 piston, the area of the labyrinth seal.
(~T s
\/ 13 MR. RUBENSTEIN: There is something I'm missing. ,
14 What initiates the lifting of the tube away from the ball nut?
15 MR. KOBSA: Well, this is not held down there at all 16 and the water is free to pass within the nut and into the 17 hollow piston. This is not by any means a full metal to metal 18 contact.
19 MR. RUBENSTEIN: Okay. It looks like it from the 20 drawing.
21 And do you run the electric drive up then? The 22 hydraulic scrams it in.
23 MR. KOBSA: And besides the hydraulic scram action, 24 following the scram or immediately with the scram there is a 25 nut run-in that is automatically triggered so the electric Heritage Reporting Corporation (202) 628-4888 l
126 1 motor backs it up and runs the nut in.
2 Before that, when the rod scrams in there or latches 3 on the piston which opens into the guide tube in its fully 4 inserted position, in fact they would open into positions 5 intermediate also if there was at any time a separation or a 6 partial scram.
7 MR. MICHELSON: Is that nut trying to spin as the 8 hydraulic system is injecting the rod?
9 MR. KOBSA: It doesn't spin. It prevents it from 10 spinning because a roller on it rides in a groove on a guide 11 tube.
12 MR. MICHELSON: Oh, it's got a groove. Okay.
() 13 MR. KOBSA: So it doesn't spin but the spin, the ball 14 screw turns and it spins. So it does spin.
15 MR. MICHELSON: What latches in then?
l l 16 MR. KOBSA: The scram is all over with in a second L 17 and a half.
18 MR. MICHELSON: Right. I understand that.
l 19 MR. KOBSAt And the rod starts to move.
20 KR. MICHELSON: What disconnects the mechanical drive 21 from the piston while it's trying to inject under hydraulic 22 pressure?
' 23 MR. KOBSA: What disconnects it?
24 MR. MICHELSON: Yes. That's that little latch in O 25 there? '
Heritage Reporting Corporation (202) 628-4888
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() 127 1 MR. EBERSOLE: The black part goes up.
2 MR. KOBSA: This dark part is the hollow piston and 3 it goes up as a resitlt of the net force of the hydraulic fluid 4 on that side.
5 MR. MICHELSON: The ball there is going up with it in 6 the groove?
7 MR. KOBSA: There is a ball that go with it in the 8 groove, yes, that does prevent the piston from turning 9 separately from the one on the nut bolt.
10 MR. MICHELSON: I got you.
11 MR. EBERSOLL: The exhaust water goes past the 12 labyrinth seal on into the vessel?
r'
\ )T 13 MR. KOBSA: Yes.
14 MR. EBERSOLE: And that, now, when the vessel is 15 under full pressure, you have more than adequate over pressure 16 to drive it on against the internal pressure?
17 MR. KOBSA: Yes. And of course it is that. It puts 18 it by brute price. There is no dumping of the top side.
19 MR. EBERSOLE: And when there is ao pressure at all 20 like in a LOCA it will just go in a little faster?
21 MR. KOBSA: Absolutely. Quite a bit faster. In 22 fact, about twice as fast.
23 MR. EBERSOLE: Right. And do you have new higher 24 pressure in fact to do this? You used not to have enough 25 pressure or volume to do it. You've increased the pressure, Heritage Reporting Corporation (202) 628-4888
() 128 1 haven't you, in the accumulator?
2 MR. KOBSA: The pressure in the accumulators is 3 about, is almost 2,000 psi, yes, that's a little higher than
- 4 before.
5 The volume of the accumulators is quite a bit bigger 6 and of course that's because also we have two drives per 7 accumulator on this.
8 MR. MICHELSON: You no longer depend at all on 9 reactor pressure to help the rods go in?
10 MR. KOBSA: That's correct.
11 MR. MICHELSON: But you have to have a higher 12 pressure than reactor pressure to drive this in, and you said a
() 13 little over 2,000 I thought. But reactor pressure can -- okay.
14 It would have to --
15 (Simultaneous voices) 16 MR. EBERSOLE: Does ATWS pressures lock out the 17 hydraulic insertion?
18 MR. KOBSA: No, they do not. We have a criteria for 19 scram at normal operating pressure and another slower response 20 time requirement for pressures that include both the pressure 21 transients and the sizing, the refill sizing criteria.
22 MR. EBERSOLE: And the ATWS pressure which is I think 23 higher than normal.
24 MR. KOBSA: Yes.
f O 25 MR. EBERSOLE: Are the warm drives, do they have more l
l Heritage Reporting Corporation
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I thrust than the hydraulic drives to override the internal high 2 pressure?
3 MR. KOBSA: The screw drive?
4 MR. EBERSOLE: Yes-.
5 MR. KOBSA: Does not have as much thrust as the 6 hydraulic pressure, no.
7 MR. EBERSOLE: So you would then overcome the high 8 ATWS pressure with the hydraulic pressure?
9 MR. KOBSA: As far as being able to overcome 10 pressure, the screw drive doesn't overcome the pressure, it's 11 just lifting the weight.
12 MR. EBERSOLE: What I'm trying to say is this. I
() 13 would hate to see that an ATWS becomes autocatalytic and locks 14 the rods out.
15 MR. KOBSA: No, it does not. It's a sort of response 16 of scram because the higher reactor pressure --
17 MR. EBERSOLE: That depends on the calculation of the 18 ATWS pressure.
19 MR. SAWYER: Which is less than --
20 MR. EBERSOLE: Oh, now it's gone down.
21 MR. SAWYER: Yes.
22 MR. EBERSOLE: How much is it? '
23 MR. SAWYER: On some of the earlier BWRs that are in 24 service there wasn't as much release capacity as we've had on
( 25 other models.
l Heritage Reporting Corporation (202) 628-4888
4 O 130 1 MR. SAWYER: So now we've got a new calculation on 2 pressures under ATWS conditions?
3 MR. SAWYER: That's right.
4 MR. MICHELSON: Let me ask if you had no hydraulic 5 pressure and were depending upon the warm drives, what reactor 6 pressure would you reach before the warm drive could no longer 7 provide an adequate thrust?
8 MR. KOBSA: The warm drive isn't influenced by 9 pressure.
10 MR. MICHELSON: All I see is a piston equal to the li cross section of your labyrinth seal at the top. If the 12 reactor is at much higher pressure, it's thrust would go up.
13 MR, KOBSA: This communicates directly with the 14 reactor. And the system, the hydraulic system of the drive 15 provides a higher pressure than the reactor to inject the rod.
16 But the warm does not work against pressure. The pressure 17 fills the entire drive and the warm only -- the piston is 18 lifted by the nut and any differential pressure over the system 19 is seen by the nut, but from a standpoint of just running a rod 20 in at a relatively low velocity, that doesn't amount to much 21 pressure.
22 MR. MICHELSON: It'll equalize on both sides?
23 MR. KOBSA: Yes.
24 MR. MICHELSON: I got you.
25 MR. KOBSA: This is completely full of reactor or Heritage Reporting Corporation (202) 628-4888
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() 131 1 pressurized water that's throwing pressure to the reactor.
2 MR. MICHELSON: Pushes it through the friction 3 leakage.
4 MR. KOBSA: right.
5 MR. MICHELSON: And a check valve keeps it from 6 coming back.
7 MR. KOBSA: The check valve keeps it from coming back 8 if this line breaks or there was no pressure in the system.
9 MR. RUBENSTEIN: I have another simple question. I 10 drive the ball nut up and drive the rod in.
11 MR. KOBSA: Normal position.
12 MR. RUBENSTEIN: Right. Now, I start driving out by
() 13 driving the ball nut down.
14 MR. KOBSA: right.
l
[ 15 MR. RUBENSTEIN: What brings the rod down?
16 MR. KOBSA: Strictly weight.
17 MR. RUBENSTEIN: Just the weight of the rod.
18 MR. KOBSA: Just the weight of the rod and the moving 19 piston.
20 Let me at that point out then that how this is 21 sensed, the separation is sensed by the instrumentation in the 22 drive and that precludes the rod drop.
23 First of all, the rod is coupled to the drive with a 24 bayonet type coupling. It retains the spring fingers of the 7_
V 25 drive coupling that we've always had to take up the clearance Heritage Reporting Corporation (202) 628-4888
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in the coupling so that when the drive is scrammed or the drive 1-2 is stopped at the end of the scram and takes up that clearance-3 you don't get a big impact blow.
4 But it is also a bayonet that is engaged by a quarter 5 turn rotation of tha blade relative to the drive and that's why 6 therollersareonthEhollowpistonsothatitcannotturn 7 once the drive is installed, the control rod cannon turn once 8 it is installed because it passes through the fuel support 9 castings and between the fuel itself and so once it is 10 installed it cannot turn.
, 11 So once coupled there can be no uncoupling because 12 you can't get relative rotation and it takes relative rotation
() 13 of a coupling to take it apart, to get it into a position where 14 that bayonet can come upon it.
15 So that's one picture, that the hollow piston and the 16 blade will stay together.
17 The other thing is that the entire control ' rod, 18 hollow piston and spindle and ball nut, that whole chain is 19 supported on a spring load and p.latform from below, so it can 20 travel up and down.
21 And there is a magnet built into that platform and 22 the position of that magnet is sensed with switches, magnetic 23 switches from the outside to sense the position of the 24 platform. So that's just a scale really that weighs this.
25 And it's adjusted so that if the weight, if the Heritage Reporting Corporation (202) 628-4888
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133 1 combined weight is not sufficient to depress the platform, then 2 two redundant Class 1-E switches signal that separation, that 3 relief of weight that would be caused if for any reason the 4 control rod and piston or just the control rod for that matter, 5 would become separated from the nut. Then that would be that' 6 much less weight on the spindle, that much of a change in 7 position of the platform and would signal that the control rod 8 is not following.
9 At that time, you get a rod block so that you can't 10 pull the nut out from under the control rod and that 11 prevents --
12 MR. RUBENSTEIN: But how do I get the rod out?
13 Suppose the rod is stuck in the shell somehow and we get the 14 separation, we get the rod block, but what ability do I have 15 now to pull that rod out?
16 MR. KOBSA: To pull it? You can't.
17 MR. RUBENSTEIN: You can't.
18 MR. K0BSA: You can only insert it. So you'd have to 19 insert and block it in.
20 MR. EBERSOLE: I'm getting a little bit of a 21 disturbing impression that that's a leaky piston, because you 22 told me --
23 MR. KOBSA: This is a leaky piston?
24 MR. EBERSOLE: Yes, the bottom, the black part.
)
25 Right there. That's a leaky piston, because you said it Heritage Reporting Corporation (202) 628-4888
i 134 1 -equalized if you went up on the mechanical warm' drive.
2 MR. KOBSA: .No.
3 MR. EBERSOLE: Let me just ask what I was going to 4 ask, o
! 5 MR. KOBSA: Okay.
6 MR EBERSOLE: When you execute a ceram function, 7 what keeps the rod in the scram position without the assistance 8 of a screw drive?
9 MR. KOBSA: Two things keep it there. One is the ,
P 10 hydraulic pressure from the system. The other is that there 11 are latches. In the final analysis, the answer is latches.
12 MR. EBERSOLE: Okay.
) 13 MR. KOBSA: There are latches that open on this.
14 MR. EBERSOLE: I didn't hear the latches.
15 MR. KOBSA: Engage in the guide tube and that ,
16 prevents the rod from backing down.
17 MR. EBERSOLE: I was anticipating the leaky piston to 18 let it come down.
I 19 MR. KOBSA: No. It holds it up. There are spring 20 latches that hold it there until the nut follows up and picks
- 21 up the supoort.
22 MR. EBERSOLE: Okay. I didn't see the latches.
23 MR. KOBSA: They are not illustrated and I don't even
-24 see --
0 25 MR. EBERSOLE: But it is a leaky piston isn't it?
4 Heritage Reporting Corporation (202) 628-4888 i
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n 135 s ,, - . 1 MR. MICHELSON: Yes. It has to be, h ~2 MR. KOBSA: 'Yes, it's relatively leaky compared to 3 the piston seals which we have~on the lock and piston drive, 4 but this labyrinth seal is the seal that restricts the leakage.
5 MR. EBERSOLE: It has to be a leaky piston to let the
-6 warm drive work, doesn't it?
7 MR. KCBSA: Well, it can't be absolutely 8- hydraulically locked, you're right, or else the warm drive 9 wouldn't work. That's right.
10 MR. MICHELSON: Actually, that scram piston is not a 11 piston in the sense it rings on metal. It's the area' equal to 12 the labyrinth seal that counts.
13 MR. EBERSOLE: It's a labyrinth seal.
14 MR. KOBSA: Yes, there's no seal down here, no, none 15 at all. There is no seal around this at all.
16 MR. EBERSOLE: So you have a good bit of water 17 squirting past that end of the vessel while you execute a 18 scram.
19 MR. KOPh\s Only what goes through the labyrinth, 20 that does the sealing and this OD of the piston here is free, 21 relatively free, you know it's like about a sixteenth of an 22 inch I would say on the side here.
23 MR. EBERSOLE: Wait a minute. The difference of 24 pressure across the labyrinth has to be lower than that across
() 25 the piston or it wouldn't ever move.
Heritage Reporting Corporation (202) 628-4888
1 MR. KOBSA: The labyrinth creates the pressure.
2 Imagine on the Fourth of July when you put a tin can and 3 another tin can and a firecracker and let it off and there goes 4 the other tin can out the end.
5 MR. EBERSOLE: But the exhaust side has got to lose 6 water faster thaa the input side so you get some sort of load 7 on the piston.
8 MR. KOBSA: It's true that the water in the
! 9 displacement of the control rod has to --
10 . MR. EBERSOLE: ' lou better let it leak out faster than 11 it's coming in or you'll be in trouble, on the back aide.
12 (Continued on Page 137)
(:) 14 15 16 17 18 19 20 21 22 23 24
() 25 Heritage Reporting Corporation (202) 628-4888
137 1 MR. KOBSA: When the piston moves up, the water from 2 that cube there has to make up for that displacement.
3 MR. EBERSOLE: I don't know.
4 MR. KOBSA: And so it is a displacement of the piston 5 itsel f, that is being provided by the accumulator.
6 MR. EBER' SOLE : All I am saying is the relar'tr 7 leakage rate is such that the labyrinth must leak faster than 8 the piston.
9 MR. KOBSA: No, no. You see, the water above - this 10 water above this part here ir free to communicate not through a 11 seal at all but through a much, much freer path to communicate 12 to the underside of the nut.
i 13 MR. EBERSOLE: All right.
MR. KOBSA: The seal is on this diameter of the
(]) 14 l 15 hauling, and that is the only seal. That is the only I 16 restricture.
17 MR. EBERSOLE: That is putting it up as a smaller 18 diameter sectional area.
19 MR. KOBSA: Right.
20 MR. EBERSOLE: Okay. I got it.
21 MR. KOBSA: It is a smaller area. It is not really a 22 piston.
23 Two separate sources, two separate paths, cause the 24 run in of the electric by the electric motor. One of them 25 there is the automatic nut run-up and another el'ectric run-in, O
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V 1 it is called. One comes from one path through the reactor trip 2 system and the other through the rod control system.
3 MR. MICHELSON: We will have to speed up a little 4 better. We are cutting into the lunch hour, now.
5 MR. KOBSA: We are using the committee's question 6 time.
7 MR. MICHELSON: We are not going to eat if you ask 8 him any more questions.
9 MR. KOBSA: Okay. As mentioned, there is no scram 10 discharge volume, then everything is in and that minimizes any 11 contamination of the drive. In fact, the reactor water does 12 not ever come down into the drive and doesn't go out into the 13 scram dump.
( ) 14 The electric motor control does provide for 15 maneuverability and complements load flow and load following, 16 and as you recognized, this is conducive to gang grot 17 operation. It is intended that this be operated in as'many as 18 26 drives can be ganged together for electrical positioning.
19 The materials of construction here are all proven 20 corrosion resistant materials. The main members of the -- the 21 pressure boundary of this piston tube, the hollow piston tube 22 and the outer tube, are XM-19 and the shaft is 17-4 23 precipitation hardened. Those are the main materials of the 24 drive.
25 The coupling spud is Inconel 750.
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139 1 The hydraulic 7ystem that does this in the way of
}
2 scram is illuctrated on the next chart. The control rod 3 charging pump which is a non-ecsential pump and system for 4 filling the accumulators provides after prefilling the gas 5 bottle and cylinder with precharge of nitrogen, it builds us 6 the pressure in the water side to a pressure between 1850 and 7 2000 and that then is released through a scram valve.
8 There are one line per drive. This illustrates the 9 two drives per accumulator that are serviced by the one 10 accumulator, and the purge line that provides the purge flow, 11 as I said, to keep the drive clean and not necessarily cool.
12 There is no requirement for cooling.
13 MR. EBERSOLE: You still have to provide a device to
()14 assure you will get a broke, and total loss of air pressure to 15 inake it work? Instead of a gradual? You know, you had trouble 16 with that.
17 A gradual reduction of air pressure of control air 18 will cause the lockup, and I think you had to put some kind of 19 a modulator in the system to ensure --
20 MR. KOBSA: On the scram value solenoids?
21 MR. EBERSOLE: Yes. When you have got a slow leakage 22 in pressure, they got to some 3 7termediate point and they 23 fluttered and they wouldn't work. I don't remember the details 24 of this.
25 MR. KOBSA: I am not aware of that, particularly.
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(} 1 MR. MICHELSON: Yes, that is well established. That
]
2 is a part of this analysis that we would expect to see, i l
3 eventually. j 4 MR. KOBSA: Okay. The next slide -- to illustrate, 5 and rea.lly the time to da that -- but rather than the 6 schematic, and I have tried to describe that in words, the 7 drive mechanism really being a pipe within a pipe and the 8 labyrinth seal.
9 I didn't mention particularly the thing that stops 10 the drive at the end of this stroke is a stack of Bellevil.le 11 springs. So it is spring buffer that stops the drive at the 12 end of the stroke, and that is a very reliable buffer but we do 13 have means to check it and test for it periodically at
() 14 shutdown.
15 MR. CORK: Irv, could we skip the next two charts.
16 MR. KOBSA: Yes.
17 MR. CORK: And skip to the one on operating 18 experience.
19 MR. KOBSA: Okay, because these are pretty much a 20 rehash of the things that we have been talking about. If you 21 when glancing at them have any questions, I will be happy to 22 try to answer them.
l 23 And as far, then, as the experience is concerned this 24 is a drive which I think in Dan's remarks to begin with was 25 mentioned that it was originally, the concept was originally a Heritage Reporting Corporation (202) 628-4888 l -
i 141 (3 1 design developed by GE long ago, and then TWU picked it up and ,
\/ l used it in Europe, and the Swedes have'a similar drive. !
2 3 Different only in that it rather than using a ball nut uses an 4 acme threaded nut and screw.
5 But there are nearly 2700 drives in service. They 6 have been operating for up to 15 years in the European 7 reactors, and many thousands of years of drive years of 8 experience.
9 There have been no outages resulting from these 10 drives' performance and very little maintenance is required on 11 the drive because of the relatively simply construction without 12 the seals and without the reactor water coming in to them.
13 No parts have had to have been replaced and no
( ) 14 observed damages to them.
15 And for this particular drive, there is a 16 considerable amount of testing also in its adaptation to the 17 ABWR.
18 "here was a prototype drive patterned af ter, directly 19 after the KWU drive which was built and tested having 600 20 scrams and 6700 motor driven cycles on it. Then a modification 21 of that was developed and made for ultimate use in actual 22 operating reactor testing it at LaSalle Unit 1. It is in there 23 now.
24 Prior to its installation, it had 500 scram cycles
. 25 and 63,000 stopping cycles applied to it, and that operation in Heritage Reporting Corporation (202) 628-4888
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v 1 the reactor there is continuing.
2 MR. MICHELSON: When you speak of all of this 3 experience with the drive, are you including the drive module, 4 the accumulator tanks, the scram valving, the whole bit as part 5 of that?
6 Or is that uniq'ue for this design? Do the' Germans 7 and the Swedes use the same control rod drive modules for 8 operating?
9 MR. KOBSA: No. The Swedes, at least, have a much 10 bigger accumulator. They have 12 or 15 or more drives on one 11 big accumulator tank.
12 MR. MICHELSON: The drives have been well used and 13 proved the drive units. The drive unit modules have not
( ) 14 necessarily been equally proved. Is that correct?
15 Is the accumulator tank you are proposing for the 16 ABWR the same as for the old boilers?
17 MR. KOBSA: Monkey piston?
18 MR. MICHELSON: Are you using the bladder arrangement 19 in the tank, or are you using an actual tank with a piston in 20 it? Or, just what?
21 MR. KOBSA: This is a tank with a piston and a 22 separate nitrogen storage tank.
23 MR. MICHELSON: Okay, it has got a piston ring and the 24 whole bit, then. You have gone away from the bladder
?.5 arrangement, then.
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Because the older tanks used to use an neoprene diaphragm, I believe it was, to separate the water from the l
3 gas.
4 MR. KOBSA: We have in the past both kir.ds of )
5 accumulators.
6 MR. MICHELSON: So what I would like to hear about a 7 little bit is all the experience that you have had with the 8 drive modules you are proposing to use, their reliability, 9 because there is where you are getting into the air 10 contamination questions. You get into the over-voltage 11 questions and so forth.
12 MR. CORK: Irv, aren't these accumulators the'same as
! 13 the Bietimara-6 except for the number of drives maybe per l
( ') 14 accumulator, but the accumulator and the nitrogen bottle.
15 MR. KOBSA: I am not sure, Joe, that they are 16 identical to the BWR-6.
17 MR. MICHELSON: On the next meeting, we will put it 18 on as an agenda item to get a clarification, and somebody's 19 idea of the experience that you have got with the drive units 20 and not just with the drives themselves.
21 MR. EBERSOLEi Are they using that piston as an 22 interface?
l i 23 MR. MICHELSON: Well, they are using piston-type l
24 accumulators with a piston ring arrangement and so forth, I 25 guess.
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144 1 MR. EBERSOLE: Well, they always leak and I don't g}
2 know how you would determine you had an air lock on that.
3 MR. MICHELSON: Well, once they clarify what they 4 have got they will have to tell us what kind of analysis they 5 have done of failure modes and effects on those drive units
)
6 as well as on the drive itself.
7 MR. KOBSA: That has been done in the past by sensing 8 the pressure, of course, and the air volume and conductivity 9 sensors to determine that it wasn't moisture.
10 MR. EBERSOLE: It is not the moisture that counts, it i
11 is how much volume there is at pressure.
12 MR. KOBSA: Right. So there was a stop on the piston 13 that prevented, that gave you a minimum volume.
( ) 14 MR. EBERSOLE: Yes.
15 MR. KOBSA: And then the pressure in that minimum 16 volume was sensed so you knew you had the right energy source 4 17 as long as it wasn't water.
18 MR. EBERSOLE: Yes.
19 MR. KOBSA: And the conductivity sensor was used.
20 MR. MICHELSON: In the old days of FSARs, there used 21 to be a description of the hydraulic drive unit. The whole 22 package. Pictures of it; descriptions of it; details. I don't 23 find any of that in this FSAR. I am a little surprised.
,u .
24 No reference as to where to find it or anything.
, 25 MR. EBERSOLE: Well, that doesn't mean anything if
(
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the air volume has leaked full of liquid.
MR. KOBSA: Right. That is why the conductivity j 3 sensor for determining this is not water.
4 MR. MICHELSON: I don't want to get too much into the 5 details now, but we will I think pursue this at some length 6 later on and get the information that we need with which to 'do
~
7 the review.
8 But I was a little surprised. The drawings in the 9 FSAR are totally inadequate from this viewpoint.
10 MR. WYLIE: In the operating experience, I noted you 11 say "no forced outages due to mechanical malfunction." What 12 about electrical malfunction?
13 MR. KOBSA: I am not aware of any problem with the electrical, either. The electrical motor was not the same on
( ) 14 15 the KWU drive as we are using on this drive. We have a 16 stepping motor and they had a straight induction motor, I 17 believe.
18 And I don't really know that that was intended to be 19 restrictive there. I don't think that it was, but I can't say 20 for sure.
21 MR. EBERSOLE: That business about the vulnerability 22 to air pressure as being moderately low?
23 MR. KOBSA: Yes.
24 MR. EBERSOLE: Is embodied in -- what is the general 25 study, Carl, A-177 Heritage Reporting Corporation (202) 628-4888
I 146 1 MR. KOBSA: Well, it may be in A-17. The system
)
- 2 interaction, you mean?
3' MR. EBERSOLE: No, no, no. No, the one about the 4 input to control systems.
5 MR. KOBSA: Well, that is in A-47.
6 MR. EBERSOLE: A-47. It is embodied in A-47 where 7 one was'looking at progressively degraded air systems and the 8 implication of altering the performance and safety.
9 MR. MICHELSON: You are not using a dedicated 10 compressed air for this, as near as I could tell?
11 MR. KOBSA: No.
12 MR. EBERSOLE: This is a standard air system, isn't 13 it?
( ) 14 MR. MICHELSON: Well, you couldn't tell that either 15 from reading the FSAR. Couldn't tell what the air arrangement 16 was from this chapter.
17 MR. EBERSOLE: You see, that gets into the aspect of 18 it is just a non-dedicated air system. It may be full of crud.
19 MR. KOBSA: This is a nitrogen charge from nitrogen
-20 bottles.
21 MR. MICHELSON: No, but control air -- you have to 22 have control air to control this thing. You are only using 23 nitrogen.
24 MR. KOBSA: Instrument air for the scram valve 25 control.
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147 1 MR. EBERSOLE: Precisely, instrument air.
f~}
-J 2 MR. CORK: The ventilator is charged with nitrogen.
3 MR. MICHELSON: Yes, but the control of the nitrogen, 4 the control valve is air.
5 MR. KOBSA: In the scram valve, okay, control is by 6 air.
7 MR. EBERSOLE: You all don't maintain a quality 8 ' control over that air supply.
9 MR. KOBSA: Well, we don't know, Jesse.
10 MR. CORK: We certainly specify the requirements of 11 the air.
12 MR. MICHELSON: Yes, but the realization of that in 13 the context of liquid leakage and contaminants and such that crept up all the valves simultaneously is an open issue.
(]) 14 15 MR. CORK: No. Each of those events you are 16 referring to have a fix associated with them which has been 17 carried through into this design.
18 MR. EBERSOLE: There is quality control of the air 19 purity?
20 MR. MICHELSON: Well, Jesse, we are going to get a 21 document -- some re-explanation on their failure modes and 22 effects of a control rod valve system. Because clearly it is 23 one of the most critical systems.
24 MR. KOBSA: And it will include the quality of air.
25 MR. MICHELSON: I certainly think so.
1 I
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148 f, 1 MR. KOBSA: Maintenance of air quality.
_ k..)
2 MR. MICHELSON: Yes.
3 MR. CORK: Let me confirm that I have heard mentioned
.4 that GEOs, failure mode effects analysis on the CRD system, one 5 that would include the air supply contaminants, the seal 6
- material contamination, over and under voltage, cylinder 7 voltage problems, and gradual reduction of air pressure 8 difficulties.
9 MR. MICHELSON: I think that essentially have been 10 the past problems that we would expect you to look at.
11 MR. CORK: We will provide such a report.
12 MR. MICHELSON: Yes. It doesn't have to be fancy, l 13 just has to show that it was thought through. Maybe one page will do it, I don't know. It depends on what you have to say.
( } 14 15 That wraps up your presentation, I believe.
16 MR. KOBSA: Yes, sir.
17 MR. MICHELSON: We will take our lunch break now and 18 come back at 1:15.
19 Excuse me. Do you want to comment as we go to each 20 chapter, or what would you prefer?
21 MR. KOBSA: I suppose that will be necessary. We are 22 still very early in our review process.
23 MR. MICHELSON: Yes. Would you like to hear whatever 24 we have to say on these as we go, or at the end, either way?
25 Do you prefer to do it at the end?
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MR. KODSA: We have been able to do it at the
{' 1 2 end.
Yes.
)
3 MR. MICHELSON: Okay.
. 4 MR. KOBSA: All right.
5 MR. MICHELSON: That's fine, then. We will adjourn 6 until 1:15 p.m., that will put us only 15 minutes behind.
7 (Whereupon, at 12:15 p.m., the meeting was recessed, 8 to reconvene the same day, Wednesday, June 1, 1988, at 1:15 9 p.m., in the same place.)
10 11 12 13
- O 14 15 16 17 18 l
19 20 21 I 22 23 i
24 l 25
($)
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~g 1 MR. MICHELSON: We will readjo6rn here now for a (J 2 discussion by General Electric of Chapter 5 of the SSAR.
MR. KOBSA: For the record again, I am Irv Kobsa,
~
3 4 principal engineer on the reactor design for the ABWR. And the 5 agenda for Chapter 5 would be an overview of the integrity of 6 the reactor coolant pressure boundary, the reactor vessel, and 7 components and subsystems. And in particular, in that area, I 8 will discuss I think what would be of most interest to you, the 9 new component, the reactor internal pump.
10 (Slides being shown.)
11 MR. KOBSA: The reactor coolant system is --
12 MR. MICHELSON: Before you get started, let me ask 13 you, under components and subsystems, they actually include the 14 residual heat removal I believe, or the RCIC.
(])
15 MR. KOBSA: Yes, that is true.
16 MR. MICHELSON: And there are going to be some 17 questions. I guess it is just RCIC, residual heat removal, and 18 reactor water cleanup, or parts of that.
19 You will be prepared to discuss those?
20 MR. KOBSA: I will try to answer questions if you 21 have them on those. I do not have a prepared chart or anything 22 on that.
23 MR. MICHELSON.: They seem like to be pretty well like 24 parts of things that we have seen in the past, but not 25 entirely.
O Heritage Reporting Corporation (202) 628-4888
151 rS 1 MR. KOBSA: Okay. The primary coolant boundary does b 2 consist of the reactor vessel, of course. And included in the 3 reactor coolant system are the reactor internals. The reactor 4 internal pumps are a part of that, system. The main steam, 5 safety relief, and feedwater systems out to and including the 6 isolation valves are a part of that, and they do have an 7 automatic depressurization feature.
8 The reactor water cleanup system. The high pressure 9 core flooder systems, there are two of those. And the residual 10 heat removal system, which has two modes of operation. One 11 shutdown cooling, and the other an EECS emergency core cooling 12 at low pressure, which Dr. Sawyer will be talking about both of 13 those aspects of the high pressure core flood and the low 14 pressure core flooding systems in connection with his
(])
15 presentation.
16 MR. EBERSOLE: Let me ask you a question about the 17 reactor coolant depressurization system. Historically, these 18 have been operated by DC, I believe. They must be energized to 19 hold open, on the grounds that to be closed is safe which is 20 not true.
21 MR. MICHELSON: Jesse, these are going to be 22 different kinds of valves than you are used to seeing.
23 MR. EBERSOLE: Well, that is what I was going to say.
24 To what extent are they different and better?
25 MR. MICHELSON: I think that you are going to go into
! )
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I 152 1 the relief valves a little bit, are you not?
(~')
\~s 2 MR. KOBSA: I do not have that prepared. In~ fact, we 3 have not selected. There is not a particular separate relief 4 valve.
5 MR. EBERSOLE: Well, let's say among I guess some 6' others. If you are going to vastly improve a valve which is 7 interdependent on DC control voltage under exposure to the 8 containment environment, air systems which have pollutants, and 9 a host of other things, I hope that you will have some kind of 10 warm drive in-crank of some sort in the final analysis to get 11 the damn thing open. Because to get it open in the final 12 analysis probably is just as important if not more so as to get 13 it closed. So you are caught between the two. And in the
() 14 present configuration, you lose in closing it.
15 MR. KOBSA: On the relief valves?
16 MR. EBERSOLE: On the relief valves, yes. You lean 17 strongly towards closing it. And I think that direction of 18 leaning is probably worse than other ones.
19 MR. MICHELSON: I guess that when I talked to 20 Mr. Box the other day, he did not communicate. Because I told 21 him that I was interested in the relief valves, because they 22 appeared to be unique. These are not the target rock kind of 23 animals, if I understand the report right.
l 24 MR. KOBSA: Well, we have been giving serious l
25 consideration to using solser type valves. But at the moment, I
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{) I 2
there is not a decision on that.
101. MICHELSON: I am just looking at your reports, 3 that is all.
4 MR. SAWYER: The relief valves are going to be power 5 operated. They are not pilot operated.
6 MR. MICHELSON: That is right. This is it right 7 here, I assume. It is out of your report. And we certainly 8 wanted an explanation of it, because it is unique.
9 Have you ever used these on BWR6s?
10 MR. SAWYER: Operator valves have used on almost all 11 the Ss and the 6s. The power is there.
12 MR. MICHELSON: The big pneumatic operator hanging on L 13 the side, it is more like a standard manual valve with an t
() 14 operator on the handle.
15 MR. EBERSOLE: Where is the valve that opens and 16 shuts the air supply, is it outdoors I hope?
17 MR. MICHELSON: They did not even discuss it.
18 MR. EBERSOLE: If you hang it inside, you are about 19 where you started.
20 MR. MICHELSON: That is right.
21 MR. EBERSOLE: So I do not know what we have got. By 22 the way, I have not been getting any papers.
l 23 MR. MICHELSON: Did you not get that?
24 MR. EBERSOLE: Well, I have not been home in a week.
25 MR. SAWYER: What section is this?
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154 1 MR. MICHELSON: This is Section 5.2.
2 MR. EBERSOLE: I will take inventory. I have not get 3 that stuff.
4 MR. MICHELSON: We can get a copy of this made.
5 Because we are going to discuss this valve a little bit today.
6 Or I guess that we will have to delay it until the next 7 meeting.
8 MR. KOBSA: I am not really prepared to give you any 9 details on the valve itself myself.
10 MR. MICHELSON: This is quite a bit different than 11 the old ABWRs at least.
12 MR. EBERSOLE: Well, there were a number of 13 applications and designs where it was thought that a selected 14 direction of operation was the preferred one. And it turns out
(])
15 that lots of them have a bi-directional responsibility which is 16 not addressed in the design that we have today. I think that 17 whole topic of which way is safe has to be investigated. And 18 in fact, you might find that neither way is safe, that you have l
19 got to go both ways.
20 MR. MICHELSON: This is not designed for POP action 21 either. This does not have the 50 pound hang-up that the old l 22 ones had, but it has got'a few other unique questions.
l 23 MR. EBERSOLE: The 50 pound hang-up goes back to that 24 80 pound question that we had awhile ago which negates the 50
' 25 pounds.
C)
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(~y 1 MR. MICHELSON: This was an old manual-operated valve L'
2 with an operator hanging on the side.
3 MR. EBERSOLE: Right.
4 MR. SAWYER: These valves are manufactured by
~
5 Crosby Company.
6 MR. EBERSOLE: That is a vast improvement.
7 MR. SAWYER: These valves cannot reclose with 8 container back-fire.
9 MR. EBERSOLE: That is right.
10 MR. MICHELSON: And that is a nice advantage. There 11 are some questions though about the rate at which that thing 12 will consume air in the accumulators and so forth. Because 13 they are being held open by the accumulator air. And if they
() 14 leak much, you do get many cycles out of one.
15 MR. SAWYER: Well, let's take that one step at a 16 time. I am sorry for going off the end here. I guess that in
- 17. preparing this presentation, we thought that you wanted to hear 18 about things that are very new. Whereas this area of the 19 design is similar to the GE SSAR design, so we did not 20 concentrato on it. We can prepare information for later if you 21 want.
22 There are eighteen valves. Eight of them have this 23 so-called ADS function. The ADS function has accumulators.
24 Actually, they all have. The ADS function has accumulators.
25 And the ADS function also has a dedicated nitrogen supply to Heritage Reporting Corporation (202) 628-4888
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r3 1 hold the valve open. And the location of the nitrogen supply
(_)
2 is in a location where I think that our' commitment is for a i 3 seven day response after an accident. l l
4 MR. MICHELSON: For how many cycles or how long can I 5 keep those?
6 101. SAWYER: The accumulator itself without the need 7 to call on the back-up nitrogen that is in the bottles is I 8 think about five POPS per valve.
9 MR. MICHELSON: These have to have air to keep them 10 open.
11 MR. SAWYER: They need air to keep them open. The 12 accumulators that are attached to each of the valves for ADS 13 service are designed to have the valve open and close. It is 14 the size of the accumulators.
(])
15 MR. MICHELSON: How much leakage will some of these 16 pistons have, air leakage?
17 MR. SAWYER: I do not recall. Of course, that is 18 part of the qualification program for the valves and in
[
l 19 accepting the valves, proving that they have that capabilities.
20 MR. MICHELSON: It is also part of saying that there 21 are five cycles, and if the five cycles are spread over eight 22 days.
23 MR. SAWYER: No, no. The accumulator supply is only 24 for the short-term. The long-term supp1'y is provided by means 25 of the nitrogen bottles which will hold it open indefinitely, O
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157 1 given that we have access every week.
2 MR. MICHELSON: I am just not a very good reader I 3 know, and I had a lot to read, but was that explained in the 4 SSAR?
5 MR. SAWYER: I believe so. While you are going on to 6 the other discussion, why do I not go and see if I can find it 7 for you.
8 MR. MICHELSON: Yes, find out the section that I 9 should have read. It should be in Chapter 5.2.
10 MR. SAWYER: It may be in 6.3. That is why I have 11 got to look, because that is the EECS.
12 MR. MICHELSON: Maybe it was, maybe I missed it.
13 MR. EBERSOLE: This is combined safety and relief.
() 14 MR. SAWYER: It is combined safety and relief. It 15 will release itself on spring alone. And in addition, it can 16 be forced to open with air power.
17 MR. EBERSOLE: Now where is the trigger that opens 18 and closes the valve which supplies the air?
19 MR. SAWYER: The trigger is the DC solenoid.- You are 20 correct, Jesse.
21 MR. EBERSOLE: Is it outdoors somewhere?
22 MR. SAWYER: No. It is one of the few things that is 23 located inside the drywell.
24 MR. EBERSOLE: Why do you not get that outside of the
- 25 drywell and put it where you can fix it; why does it have to be
(
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158 1 deliberately subjected to environmental challenges; why do you 2 not bring the air line in from it and let that be out like you 3 used to have things on the old MARC-Is?
4 MR. SAWYER: I do not have an answer to that. I will 5 have to get back to you.
6 MR. EBERSOLE: It just seems asinine to invite 7 trouble by putting sensitive devices in a hostile environment.
8 MR. SAWYER: That is a good question, and I do not 9 have the answer. We will have to get back to you.
10 MR. MICHELSON: They are inside the containment in 11 MARC-I.
12 MR. SAWYER: On all of our designs, the solenoids i 13 which trigger the relief operation.
14 MR. EBERSOLE: That is true. They are one of the few
(])
15 devices. Most of them have impulse lines outdoors, you know.
16 That is true, you are right.
17 MR. MICHELSON: We will get to the details of it 18 later I guess. But I had asked to hear about this, because it l 19 was a little different. I am not as well-acquainted, and I did 20 not remember. I thought that the BWR6s were still using the I
21 trigger.
22 MR. EBERSOLE: I do not think an air supply line has 23 an environmental vulnerability to compare it to a solenoid 1
24 valve.
l l 25 MR. MICHELSON: Okay. Why do not proceed now that we
(
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'S 1 have got that confusion straightened out.
(J 2 MR. KOBSA: The introduction of the reactor internal 3 pumps does result in.the elimination of the large recirculation 4 nozzles below the core, and improves core flooding and safety 5 performance in that respect. The internal steam separation is 6 of the BWR6 type.
7 This large reactor pressure vessel diameter moves the 8 reactor vessel even further away from the core than it has been 9 in the past BWRs. This reduces the affluents to very low 10 values. As a matter of fact, the affluents at forty years at 11 the surface, the maximum is calculated to be four times 10 to 12 the 17th. And thus, it provides for even larger margins than 13 we have had in the past with the BWRs.
14 The elimination of the external recirculation piping
(])
15 and valves makes for a containment configuration which can bc 16 better rationalized and simplified around the reactor. And of 17 course, it eliminates the need for in-service inspections of 18 those systems.
19 The internal pumps are variable speed to adjust the 20 flow, and from that the power of the reactor without that 21 motion. And the tn=ernal pump inertia provides for a slow 22 enough coastdown to keep the fuel within the its limits.
23 As mentioned before, the flow restrictors are in the 24 RPV steam outlet nt@zle. And in doing that, they do reduce the 25 affects of the loss of coolant assumed accident on both the O
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1
~
1 reactor internals and on the containment.
(m)3 2 The reactor coolant pressure boundary does comply 3 with 10 CFR 50 end the code requirements, including code case 4 approval of not only Class 1 as in the past, but Class 1, 2, 5 and 3 components. And the over-protection prevention does 6 conform to 10 CFR 50, Appendix A.
7 MR. EBERSOLE: Let me ask a question about that.
8 MR. KOBSA: Yes.
9 MR. EBERSOLE: There is an interesting history of 10 that. At one time, your reactors I think, I forget where you 11 quit doing this, but they had relief protection, assuming that 12 the rods that you stated stayed at steady full power, that you 13 could relieve at full steam. And then along about 1967 or (3 14 thereabouts, I recall that you cut back to 66 percent. Because
%/
15 the mechanical code permitted you to invoke the theory that you 16 could chop the power down by scram.ning it.
17 And then in that regime, you went down to 66 percent 18 of full steam flow. And then I understood you to come up to 19 100 percent again, which permits at least the notion that you 20 could have full bypass.
21 What is it now, what is the percent, what is the 22 percent of full forward steam relieve valves?
23 MR. KOBSA: I think that we have to ask Craig to 24 answer that.
25 MR. EBERSOLE: Do you still use 100 percent?
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161 1 MR. KOBSA: I cannot answer that question for you, (s'N
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2 but I think that Dr. Sawyer could. He just stepped out for a 3 moment, I guess.
4 MR. EBERSOLE: That gave rise to a lot of questions, 5 among them being ATWS.
6 MR. KOBSA: Excuse me, whidh valves?
7 MR. EBERSOLE: The safety relief valves. They were 8 reduced to 66 percent in 1966. And part of that time, they had 9 been 100 percent.
10 MR. MICHELSON: It is not clear.
11 MR. EBERSOLE: What happened was that it was an L
12 interesting interface between the ANS code and the nuclear 13 folks, where the nuclear folks convinced them of zero power.
I
() 14 MR. MICHELSON: They are still using electronic 15 protection as part of the SSAR, but they did not say what l 16 fraction was mechanical.
17 MR. CORK: If you are talking an isolation transient, 18 the safety relief valves can relieve full reactor pressure.
19 MR. MICHELSON: You misunderstand the problem. If I 20 run the reactor at full power and I isolate and do not scram 21 the reactor, just hold it at full power, can I relieve full
! 22 power through the relief valves.
23 MR. CORK: I see.
- .24 MR. MICHELSON
- I do not think you can, i
25 MR. EBERSOLE: It was prior to 1966.
l _
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162 1 im. CORK: He is talking about continued operation, 2 continuing into the pool and you are raising the temperaturo.
3 It cannot persist, of course.
4 MR. EBERSOLE: No, because it will get to 5 containment. But at least, it prot'ected the pressure valve 6 until you got to containment.
7 MR. MICHELSON: You see, on pressurized power 8 reactors, you can do this. You can have full power, and 9 relieve through relief valves on the secondary side of the 10 generators.
11 MR. EBERSOLE: And it was in 1966 that you made a 12 shift, and that was because of some subtle agreement between 13 ASME and'ANS I guess.
() 14 MR. CORK: Well, just let me say that given an 15 isolation transient, you have the safety relief valves, and you 16 completely depressurize the over-pressure into the pool. And 17 at the same time, you scram.
18 MR. MICHELSON: In other words, you are depending 19 upon electronics to protect the pressure boundary.
20 MR. EBERSOLE: It was not pure mechanical protection.
21 MR. KOBSA: I believe the relief valve sizing i 22 transient does include a delayed scram, not one which would 23 come with valve operation but with high flux.
24 MR. EBERSOLE: Actually, Browns Ferry has about a l
25 half dozen extra pads that are blocked off.
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'~ 1 MR. KOBSA: That is only for an instantaneous. rate of b) 2 release.
3 MR. EBERSOLE: All it does is argue that the pressure 4 boundaries are protected. It does not argue that the plant is 5 safe. Because you are going to get to containment very 6 quickly. But nevertheless, there has been some vacillation.
7 MR. MICHELSON: Why do we not proceed, and you will 8 get the answer.
9 MR. EBERSOLE: And the basis therefore.
10 MR. MICHELSOM: Why do we not wait, and he will give 11 us all of the answers.
12 MR. EBERSOLE: All right.
13 MR. MICHELSON: Go ahead.
() 14 MR. KOBSA: Okay. As you had gone into a little bit 15 before with Mr. Wade, the safety relief valves do discharge to 16 the suppression pool and limit the pressure to 110 percent to 17 the design pressure per ASME Section 3 with MSIV closure and 18 high flux scram. And therein may be the answer, that it does 19 rely on the high flux scram to maintain that pressure 20 relationship.
21 MR. EBERSOLE: In the end, that comes to a percent.
22 That is what we need to get.
23 Since you mentioned bypass, what percent bypass do 24 you have now with the MSIVs open?
25 MR. KOBSA: I do not know, I do not know.
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164 "s 1 MR. EBERSOLE: 'Well, that is a critical parameter,
)
2 the bypass percent.
3 MR. WILKENS: You are talking the bypass?
4 MR. EBERSOLE: Through the connections.
5 MR. WILKENS: Through the connections, 33.
6 MR. EBERSOLE: 33. I thought that the Japanese had 7 100.
8 MR. KOBSA: The SRVs are opening against a direct 9 spring force or. the pressure relieving and on a pneumatic 10 actuator for the relief mode, as you are aware. And the relief 11 mode of operation upon closing is cushioned, and maintains a s 12 reclosing sequence to mitigate the pressure transient and the 13 forces on the lines and in the pressure transient in the
() 14 system.
15 And the automatic depressurization, as you have heard 16 I guess before, does not make use of eight of the eighteen 17 valves for that automatic depressurization function.
18 MR. MICHELSON: When you open on a spring pressure 19 instead of pneumatic pressure, do you get a POP action out of 20 that type of valve or does it just creep open as you increase 21 pressure. This is the reason that you used to use target 22 rocks, because you did not like that creeping open because it 23 started scouring and it simmered. And so you met to target 24 rock type valves. And now you have made a full circle back to 25 a standard spring loader relief for safety.
1
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165 l 1
1 So if you open on pressure alone, you do not get a !
(} l 2 POP action, or is there something that I am missing? l 3 MR. EBERSOLE: Yes, you do. You get the impulse. !
4 MR. MICHELSON: I want to get their answer. I do not 5 think that this works that way, but I am not sure.
6 MR. KOBSA: I am not sure of this.
7 MR. MICHELSON: Well, the gentleman who is going to 8 come back perhaps can answer that one, too.
9 (Pause.)
10 MR. MICHELSON: There are two questions. The first 11 one is what percent of reactor power can be handled by the 12 safety relief valves that are on the vessel; in other words, if 13 you get a main steam isolation without a scram, what percent of
(_s) 14 that power could be handled by the through-put of the valves?
15 MR. SAWYER: The answer is 100 percent.
16 MR. MICHELSON: You have got 100 percent relief?
17 MR. SAWYER: Yes, but not rate of pressure. We have 18 got 100 percent relief at around 1200 pounds, as I remember.
19 MR. MICHELSON: The SSAR then is a little deceptive, 20 because it talks about requiring electronic protection as well 21 as mechanical protection, which infers that you have got to 22 have a scram to get full power protection.
23 MR *. SAWYER: What section?
l 24 MR. MICHELSON: Well, I will have to find it.
l 25 MR. SAWYER: In Section 15, we talk about l
l
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166 1 over-pressure analysis. And in this section, we talk about
(]}
2 over-pressure analysis.
3 MR. MICHELSON: Well, it is in Section 5, I believe.
4 MR. SAWYER: And my memory is that the statement in 5 there says 80 percent relief capacity, but that is at nominal 6 pressure. At 1200 pounds, it amounts to 100 percent.
7 MR. MICHELSON: So if you are operating at design 8 pressure.
9 MR. SAWYER: If you are operating at operating 10 pressure, it is around 80 percent power.
11 MR. MICHELSON: 80 percent power.
12 MR. SAWYER: But below design pressure, which is 13 1250, you can carry it 100 percent.
() 14 MR. MICHELSON: I will find the words in a little 15 bit.
16 MR. SAWYER: Okay. ,
17 MR. MICHELSON: What the conflict is.
18 The other aspect of the problem was is there any POP 19 action to the relief valves at all?
20 MR. SAWYER: What do you mean POP action?
21 MR. MICHELSON: If you slowly raise the pressure, 22 when the valve proceeds to open, does it pop open. The target 23 rocks do. But spring loaded safety generally cannot.
24 MR. EBERSOLE: Come on. Iron horses do that.
25 MR. MICHELSON: Well, let me get the answer.
Heritage Reporting Corporation (202) 628-4888 s
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167 l' MR. SAWYER: You are talking about the spring action O(~N -
2 or the power action?
3 MR. MICHELSON: No, the spring action only. In other
-4 words, they are opening on pressure instead opening on the 5 actual.
6 MR. EBERSOLE: You have to tell us yes.
7 MR. MICHELSON: Well, Jesse. If you do not'know, we 8 will get the answer next time.
9 MR. SAWYER: I do not know, but I can get the~ answer 10 for you.
11 MR. MICHELSON: We will get it for the next meeting.
12 MR. EBERSOLE: Well, you want the whole picture. You 13 want blow-down, too, percent blow-down.
() 14 MR. MICHELSON: Well, he just gave us the capability.
15 MR. EBERSOLE: No, no. When it pops, it relievas, 16 and then it goes down to some pressure below normal.
17 MR. MICHELSON: These are not target rock. I assume 18 that these do not go down to say a 50 pound threshold and then 19 reclose automatically.
20 MR. SAWYER: You are talking about on spring 21 pressure?
22 MR. MICHELSON: Yes.
23 MR. SAWYER: There is a historesis. And as I 24 remember, I do not remember the width of the historesis on the 25 spring action. On the relief action, we have designed it for O
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{) 1 2
about 100 psi.
MR. MICHELSON: The question that you get into is if 3 you are using it as an ADS valve and you have to actuate it, 4 the target rocks could nct be held open even with the actuator 5 below about 50 pounds pressure.
6 MR. SAWYER: That is correct.
7 MR. MICHELSON: But with these, you can hold them I 8 gather all the way.
9 MR. SAWYER: You can hold them up under pressure.
10 MR. MICHELSON: Indefinitely?
11 MR. SAWYER: Indefinitely. That is because the 12 supply line pressure of the air is higher than any reasonable 13 containment pressure.
() 14 MR. MICHELSON: And that infers that there is 15 probably not a POP action when it is operating at pressures 16 along.
17 MR. SAWYER: Yes, spring alone. I have to get the 18 answer to that for you.
19 MR. MICHELSON: So what the question really gets down 20 to is do you have a simmering problem then on this type of 21 valve?
22 MR. SAWYER: We would in any event, because the set 23 point for the safety action, as I remember, is over 1200 24 pounds. As opposed to the target rocks, where the set point, 25 because it had a combined operation as a pilot valve, the set Heritage Reporting Corporation (202) 628-4888
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{} 1 point had to be the set point for relief action also, which means that it was only within a 100 pounds of normal operation. l 2
3 That is where the simmering problem came from.
4 MR. EBERSOLE: Well, these things, if they operate as 5 I expect they must, the initial is by static pressure. And 6 then that emits steam flow, and you get an impulse throw on a 7 larger piston diameter, and that throws the pool open. And it 8 then proceeds to go from the set point pressure to open to some 9 number of pounds below normal pressure.
10 MR. SAWYER: You mean then it recloses?
11 MR. EBERSOLE: Yes.
12 MR. SAWYER: That is correct.
13 MR. EBERSOLE: It is called the blow-down percentage.
() 14 MR. SAWYER: Yes, and I do not know what it is for 15 the spring action.
16 MR. EBERSOLE: But it has to snap open. There is no 17 other way. It cannot simmer.
18 MR. MICHELSON: Safety valves do.
19 MR. EBERSOLE: Not steam driven. Water valves maybe.
20 Water, but not steam valves.
21 MR. MICHELSON: Let's proceed.
22 MR. SAWYER: Do you have other questions?
23 MR. MICH3LSON: No, that was it.
24 MR. KOBSA: In continuing with the material. The 25 reactor coolant pressure boundary material and the water Heritage Reporting Corporation (202) 628-4888
170 T 1 chemistry. ABWR materials selection, we looked at "musts". To
(./
2 have successful reactor operating experience, and be fully 3 qualified to avoid IGSCC problems, and must include the known 4 metallurgical improvements that have been recognized through 5 testing.
6 Environment'el controls, that is the water, 7 particularly the water quality, is integrated with the material I
l 8 selection. And for the ABWR, the EPRI BWR owners water 9 chemistry guidelines are intended to be followed with hydrogen 10 water chemistry available to provide additional margin agcinst 11 dust corrosion if it is used. It is not an integral jart and 12 thought to be necessary as a part of the plant design.
13 And we would utilize proven materials and processes.
() 14 MR. MICHELSON: Wait a minute. Hydrogen water 15 chemistry, you said, was an optional thing then, or did I 16 misunderstand?
17 MR. KOBSA: Yes. It is not considered as a necessary 18 or a specific part of the plant design.
19 MR. MICHELSON: It would be there.
20 MR. KOBSA: It is available, and we would recommend 21 it.
22 MR. MICHELSON: Are the Japanese thinking that it is 23 necessary?
24 MR. KOBSA: No.
25 MR. SAWYER: I would characterize it as under study Heritage Reporting Corporation (202) 628-4888 l
i 171 in Japan right now. They have not decided exactly what to do.
['} 1 DR. KERR I do not understand, and I am certainly 2
3 not a materials expert, but tell me what is meant by the
. 4 statement, "To provide additional margin against SEC."
5 MR. KOBSA: Well, I think that it is like providing a 6 belt and suspenders.
7 DR. KERR: If you are not going to have SEC, why do 8 you want to have additional margin against having it. If you 9 are not sure whether you are going to have it or not, then 10 maybe it is desirable. I do not know what additional margin 11 means.
12 MR. KOBSA: Well, by the use of the materials which 13 we have selected, the low carbon grades of stainiass steel and
() 14 other materials which we have, and the stress rules that would 15 apply per our test results would indicate that there should be 16 no stress corrosion cracking. But that the use of the hydrogen 17 addition provides even greater margins in that with hydrogen 18 additions either higher stresses must be sustained before 19 cracking would be experienced. So it does provide margin, as 20 for example as measured by what stress would be applied before 21 cracking could occur.
22 DR. KERR: So it would be prudent or me to assume 23 that there is some uncertainty in your results insofar as they 24 apply to an actual operating reactor?
25 MR. KOBSA: Well, I suppose tha. you could always say Heritage Reporting Corporation (202) 628-4888 e
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"; 1 that, but from the correlation with the tests.
'v' 2 DR. KERR I mean if*I buy one of'these things and I 3 put it into operation, would you advise me that I need 4 additional margin or not?
5 MR. WILKENS: Let me add some perspective to this.
6 Certainly, the operating BWR, GE has been on record for some 7 time as recommending hydrogen water chemistry. In the case of 8 a new ABWR plant designed from the beginning with the 9 combination of materials, selections, stress controls, and 10 chemistry controls that we have built into this plant, it is a 11 much more judgmental problem than it is in the operating 12 plants.
l 13 You can present a good case here that it is not
() 14 needed. Certainly, for normal kindo of operation. But then 15 you begin to worry about things like what if you miss a crevice 16 somewhere in the design, or what if something during operation 17 the plant gets into some period of unusual chemistry 18 conditions. And under those kinds of conditions, would you not 19 be glad to have it. The people who argue for putting it on for 20 a margin use those kinds of arguments.
21 We are right now working with the Japanese on an 22 evaluation of hydrogen water chemistry. And it may be that 23 before the certification efforts are complete that we will 24 decide to put it on. Right now, as Craig said, it is under 25 study in Japan. But it is a much more of a margin, belt and Heritaae Reporting Corporation (202) 628-4888
173 1 suspender target here, than it is on the operating plants,
( }.
2 because of the other measures that we have taken.
3 MR. SAWYER: Let me add to that one other thing, too.
4 And that is that the hydrogen water chemistry does not come 5 free. One of the reasons why there is some considerations 6 against immediately agreeing'to adopt it is that there is a 7 price to be paid for increased radiation levels in that part of 8 the plant, at least based on plants that have installed it.
9 And that varies from plant to plant. I do not think that our 10 chemistry advisor is exactly sure of the rationale for why some 11 plants only go up 50 percent radiation level, and some of them
( 12 go up a factor or three.
13 Nonetheless, it is something that ic to be contended
() 14 with. That is why it is under study, as opposed to adopted, 15 because there are not all positives that you get from its use.
16 DR. KERR: But if an operator did decide to do it, he 17 is stuck with it from then on, I expect.
18 MR. SAWYER: If it becomes part of this design, well, 19 obviously you deal with extra radiation as part of the design.
20 That is for shielding and so forth in the appropriate parts of 21 the plant.
22 DR. KERR: Well, then I guess you seem to be telling 23 me that although that extra margin will be available if you 24 used the water chemistry, that you would not want to do it with 25 this design unless the shielding or whatever had taken that Heritage Reporting Corporation (202) 628-4888
174
(} 1 into account. So it does not sound to me as if it is a 2 recommendation that can be put into effect at this point.
3 MR. SAWYER: The containment design or the turbine 4 design is not developed to the point where you have to make 5 that one way or the other yet. This is the part of the plant 6 that is being flushed out right now. I think that it is a 7 cautious attitude on our part for new designs and on the part 8 of our partners in Japan.
9 Their experience with all of their second generation 10 BWRs, their latest BWR product line which have already adopted 11 virtually all of tha recommendations that we have here in terms 12 of materials control and water chemistry, has been excallent 13 without hydrogen water chemistry.
() 14 So it is not clear whether you are getting some 15 additional benefit. Each added system to the plant provides 16 added complication to the plant, too. So it is not clear that 17 you are getting something.
18 DR. KERRt I was just trying to get a feeling for how 19 confident you are of the hypothesis that the material problems 20 have been solved. Thank you.
21 MR. KOBSA: The materials which we would be using in 22 the reactor and the pressure boundary are as shown here. The 23 use for the stainless steel the 304L or 316L. You can look at 24 these as equal. Except if there is a crevice location, then 25 the 316L is what is used and given priority.
s L.)
Heritage Reporting Corporation (202) 628-4888 w -+ =- ~--- - - - - _ , , _.
a 175 1 The low carbon stainless steel. castings, and XM-19
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2 stainless steel castings are used, as is some XM-19. material.
3 Carbon steel, of course, for the lines like the feedwater, and 4- -steam, and ACCS lines. The pressure vessel itself-is of low 5 alloy high strength material,.A533 or 508 material, as has been 6 successfully used regularly in all of our vessels. And the 7 uncreviced alloy 600, the nickel chrome iron in uncreviced 8 situations.
9 I mentioned before that we do have stress rules for 10 _the application of the stainless steel in creviced locations.
11 And for the inconel 600 in either nnar viced or creviced 12 locations, we have different allowabies for those. And for the 13 XM-19 in a creviced location, we have restrictions on stresses
() 14 on those materials when there is a creviced combination.
15 The 304 and the 316 nuclear grade stainless steel,
- 16 and high toughness carbon steel, radiation resistant low alloy 17 steels, controlled composition of inconel 600 and the weld 18 metals, there we use controls on the iron content, and are 19 beginning to use niobium controls in the weld metals to be a 20 stabilized grade of inconel load metal. And restrictions on
( 21 the heat treatment of the X-750 inconel.
22 High purity stainless steel is used. There are 23 restrictions on the elements in the material, and the ratio of l
24 carbon to other alloys in the steel for high fluence I
i 25 components. Specifically, the only component which we now Heritage Reporting Corporation (202) 628-4888 L
176
() I would expect to use it on would be for the control rods 2 themselves and the carbide carrying tubes. The other materials 3 of the core structure would probably not warrant that.
4 Of course, creviced geometries have been avoided, and 5 we are avoiding them strictly in the use of inconel weld metal 6 in particular.
7 DR. KERR: Could you tell me what a creviced geometry 8 means in this context?
9 MR. KOBSA: Well, we have a particular definition, 10 where depending upon the depth of the contact that anything 11 from line contact to a very small depth, for example a 12 thousandth of an inch, if the depth of the two surfaces come l
l 13 together are one thousandth of an inch or less, just zero width
() 14 is considered a crevice. And progressively up to where you 15 have a very long gap of many inches, a quarter of an inch width 16 would be a crevice.
17 So that in places, for example, between a in-core 18 instrument and its housing or guide tube, there is an annulus 19 there which is several feet long, and it is not a crevice, 20 because of the width of it being greater than a quarter of an l
21 inch.
22 But if you have a very close contact like where screw 23 threads come together or flanges come together and you have 24 essentially line contact, then even a very short depth is 25 considered a crevice.
O l N/
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DR. KERR: Thank-you. ;
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178 3 01. KOBSA: Of course, cold work is of the austinic
{} 1 2 material is controlled and limited so as not to produce an 3 increase in hardness above a very small amount, and the 4 composition of the well metals, the ferrite control on the 5 stainless steel and the control of the well metal for the in 6 canal welding are two things that we are doing there to prevent 7 problems in those well metals.
8 As far as the reactor vessel itself is concerned, the 9 BWR has always had the benefit of having low radiation to the 10 vessel itself because of the large weather gap between the core 11 and the vessel.
12 In this reactor, the incident fluence on the surface 13 of the base metal is for 60 years, 6 times 10 to the 17th, and may greater than 1. After that, in a maximum final RTNDT,
( ) 14 b<>
15 after: 60 years of operation, as indicated here, is 50 degrees 16 Fah*enheit. Now that was computed for well metal, opposite the 17 belt line, and since we have adopted the recommendations or 18 requirements of ALWRs as defined in every guidelines of 19 eliminating the wells opposite the belt line, that is no longer 20 50 degrees, but the base metal itself would be O degrees F at 21 end of 60 year life.
22 It starts out at a minus 20 and the shift there, with 23 allowances for margin, is only the 20 degrees to bring it up to 24 0 degree F end of life time for the base metal.
25 MR. REMICK: On that assumption of the 50 degrees for Heritage Reporting Corporation (202) 628-4888
179
(} l the well metal, what was assumed on contamination?
2 MR. KOBSA: I have to refer to some notes here for 3 that chemistry.
. 4 MR. REMICK: If you can't find it, it's not that 5 important.
6 MR. MICHELSON: While you're looking for that, I can 7 ask and go back to this question of safety valve sizing. Refer
, 8 to your FSAR Section 5.2.2.1.4, which deals with safety relief
, 9 valving capacity. It says in here that the safety valve sizing 10 evaluation gives credit for operation of the scram protective 11 systems which may be tripped by either one of two sources, a 12 direct trip or a flux trip, but you never did say how much 13 credit you had to take for tripping but it says in here that
( ) 14 you include the electronic trip as well in sizing your relief 15 valves.
16 MR. SAWYER: I wasn't certain what you meant by 17 electronic trip.
18 MR. MICHELSON: That's the scram.
19 MR. SAWYER: You mean scram. The scram function.
20 MR. MICHEI. SON: Sure.
21 MR. SAWYER: Of course.
22 MR. MICHELSON: Our question was if you did not scram 23 and you were at a hundred percent power and you had closed the 24 main steam isolation valves, would the relief valves take care 25 of it?
Heritage Reporting Corporation (202) 628-4888
180 1 MR. SAWYER: Yes.
2 MR. MICHELSON: Okay, we're not talking about beyond 3 design. We're talking about within the ASME code, of course.
4 MR. SAWYER: Within the ASME code, we do not take F 5 credit for the position scram on the MSIVs, but the backup 6 scram.
7 MR. MICHELSON: Okay, you do have to scram in order 8 to have sufficient relief capacity? It's in the code?
9 MR. SAWYER: For the code analysis. That's correct.-
10 MR. MICHELSON: That was really the question.
11 MR. SAWYER: That's correct.
12 MR. MICHELSON: It used to be that you did have 100 13 percent and then you backed off to 66, and the question was what percent are you at now?
(])14 15 MR. SAWYER: Okay. Well, as I said, to give you a 16 bird's eye view of the steady steam capacity of the valves, if 17 you had the reactor at about 1200 PSI and all the relief valves 18 open, it would keep up with the steam emerging.
19 MR. MICHELSON: Okay. It wouldn't cover the 20 transient situation, but that would cover the ASME providing 21 you would creep into that condition /
22 MR. SAWYER: Yes. There is some timing involved in 23 the ASME. We don't take credit for the first gram signal we 24 might receive. We do take credit for the second one.
25 MR. MICHELSON: But you still have to take credit for Heritage Reporting Corporation (202) 628-4888
r-181
() 1 2
it working in order to stay within the ASME code, even with all your safetys open.
3 MR. SAWYER: That's correct.
4 MR. MICHELSON: Okay. Well, I think that was the 5 question.
6 DR. KERR: What is it? He would always say it'd be 7 open or the fact that it takes a lot to get them open during 8 that interim.
9 MR. MICHELSON: You've got to add the time to get 10 then open. Sure.
11 MR. SAWYER: The time delay's in there, and they 12 don't all have the same set point. They're staggered in their 13 set points.
()14 MR. MICHELSON: I thought that was the case because I 15 road this. It didn't give me good numbers but it gave me a 16 feel that you were still depending upon scram as well to stay 17 within the code.
18 MR. SAWYER: Yes. But the backup scram.
19 MR. MICHELSON: Yes. The backup scram.
20 MR. SANYER: That's no different than the process 21 we've done for coolant evaluations going back as long as I can 22 remember.
23 MR. MICHELSON: I think that's right, except maybe 24 you didn't remember back as far as I remember, 25 MR. SAWYER: Maybe very early, but at least on the 4, Heritage Reporting Corporation (202) 628-4888
182 1 5, and 6 model line, you certainly use this process.
)
2 MR. MICHELSON: It took so many valves do it 3 otherwise that you gave up on it. Browna Ferry even gave up on 4 it. It took too ma,ny valves. Does that clear up your 5 question?
6 MR. EBERSOLE: There was the matter that Frank was at 7 Browns Ferry.
8 MR. MICHELSON: Yes.
9 MR. EBERSOLE: But was not giving up. That was 10 abandonment.
11 MR. MICHELSON: GE talked the management into it.
12 Yes.
13 MR. KOBSA: But to answer your question, the l
()14 materials were considered were copper in the base metal.
15 Cepper is .05 percent. Phosphorus, .015 percent. For the well 16 metal, we had assumed, this is not, it's assumed, it's per our 17 specification which has been reviewed by material suppliers, 18 and the phosphorous for well metal was .02 percent, .020, and 19 the copper .08 for well metal.
20 MR. MICHELSON: Thank you.
21 MR. KOBSA: As far as in service examination is 22 concerned, the large vessel does have no wells in the core belt 23 line and, in fact, per the EPRI guidelines, would be made 24 completely of forged rings which would reduce about one-third 25 the length of wells that would be necessary to inspect.
I Heritage Reporting Corporation (202) 628-4888 i
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183 1 As I mentioned before, there's only one weld in the
(~ }
2 bottom head now, and that is a circle of weld between the 3 pattern of CID penetrations and pump penetrations. We have 4 developed a particular carrying machine to carry UT transducers 5 along that weld and be able to make both transverse and in line 6 with the weld examinations by direct beam and by angle beam, 7 ultrasonic examination, remotely, 8 MR. MICHELSON: How easy now will it be to do 9 internal inspection of the vessel wall which might, in some 10 cases, be what you'd like to do it. Without gen comps it looks 11 like it's not too difficult.
12 MR. KOBSA: That's right. It does make the access to 13 the inside of the vessel completely opened there in the
()14 annulus. I've always had, myself, the impression that any 15 examination that we needed to do from the inside can be done 16 with the existing, even with jet pumps, but this does, of 17 course, eliminate the partial obstruction of the jet pumps to 18 make the complete ID free for examination down to well below 19 the core.
20 So, that is an Lmproved access for any visual 21 examination or examination to be made through the cladding on 22 that side.
23 MR. MICHELSON: Now, your insulation is all mounted i
24 on the shield wall?
. 25 MR. KOBSA: Shield wall. Right.
Heritage Reporting Corporation (202) 628-4888
184 1 MR. MICHELSON: And there's how big a gap between the (v~)
2 vessel and the shield wall and the insulation?
3 MR. KOBSA: I think there's about a two foot gap.
4 MR. MICHELSON: Between insulation and the metal on 5 the vessel?
6 MR. KOBSA: Yes.
7 MR. MICHELSON: So, it's very substantial.
8 MR. KOBSA: So that the ISI inspection equipment can 9 be installed and run around the vessel without any problem at 10 all.
11 Okay. The principle new feature in the reactor is 12 the reactor internal pump. It is a wet motor seal-less design 13 with the casing furnished as a part of the pressure vessel.
( ) 14 The casing, itself, is welded into this form nozzle, 15 and I'll show you a bit more detail about that in the picture 16 in a moment. There is a solid-state adjustable frequency 17 control that sets the pump speed and then turn the re-18 circulation flow.
l l
19 There is a continuous purge of a very small flow of 20 purged water. It comes from the same source as the CRDs that 21 purge the pump and, again, the purpose for that is strictly to 22 avoid contaminat.i.on of the pumps. There's no need for a 23 cooling there, but rather just to keep the particulate matter 24 from getting into the pumps, any radioactive material from the 25 reactor cooling.
! Heritage Reporting corporation l
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185
() 1 The motor is cooled by the reactor building closed 2 cooling water system having individual heat exchangers for each 3 pump that are mounted adjacent to the pumps in the annulus just 4 inside the support pedestal for the vessel.
5 The impellers and motors are removab1'e without the 6 core straining the reactor. I can show that in just a moment 7 on.the picture. The back seating of the impeller shaft and ,
8 blow out restraint-hangers do provide redundant LOCA prevention ,
9 means on the pump casing and on the pump cover, for that 10 matter. j 11 MR. EBERSOLE: That RCC W Cooling system used to be a 12 source of trouble because it was a single system that cooled 13 the pump seals and not the motors. It cooled the atmosphere.
()14 It cooled a whole host of things inside the containment and 15 when you lost it for one reason or another it synthetically 16 produced a high containment pressure that was misinterpreted 17 for a small LOCA. Have you done anything about that? ,
i 18 MR. KOBSA: This was one the reactor water clean up .
19 system? ,
i 20 MR. EBERSOLE: The RCC W. It had a number of t
l 21 interesting interfaces with safety signals, because when you ;
f ~
l 22 lost the cooling it would err, in particular, the air pressure ,
1 23 would rise. In this case it's inerted, and you would get a t
. 1 24 false high pressure in the containment. Now, as it was way 25 back, it was a single pipe system. It had multiple pumps. I ,
O l !
1 Heritage Reporting Corporation L
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i l -
186 g1 recall it even had threaded pipes which were fixed by welding 2 somewhat late in the game.
3 Has anything been done to upgrade RVCCW to make it 4 better than that?
5 MR. WADE: Yes, it has. We have multiple divisions 6 on the system.
7 MR. EBERSOLE: Okay, good.
8 MR. MICHELSON: Is it a safety system now or still 9 non-safety?
10 MR. WADE: It's not. Parts of the reactor building 11 closed cooling water system are safety graded. Parts are not.
12 Parts inside the containment are not. They're not required 13 after a LOCA.
h 14 MR. EBERSOLE: One of the interesting offshoots from 15 that was when you lost power it would synthetically create a 16 false picture of high pressure in the containment because of 17 overheating, because this wasn't connected to the diesels. So, 18 it had a number of interesting problems that I hope have been 19 overcome.
20 MR. SAW1ER: I think this is one of the auxiliary 21 systema ve're going to cover in a later submittal.
22 MR. EBERSOLE: Good. Okay, thank you.
23 MR. KOBSA: This is a figure of the reactor internal 24 pump and this illustration does show the con 11guration of the 2b stub noz::le in the reactor vessel knuckle section. This is dk Heritage Reporting Corporation
$h (202) 628-4888
187 that single circumferential well between the cap, which is one
(]) I 2 large forging, and this knuckle region, which is another large 3 forging.
4 It goes from this circumferential weld here to this 5 circumferential well tying to the straight lowest shell course 6 in the vessel. That single piece ring forging is made and then 7 this contour is machined into it so there is no weld in this 8 geometry which forms the penetration into which the pump 9 casing, which is also a low alloy steel forging of the same 10 material as the pressure vessel itself, is inserted and welded 11 here in a weld that is the one weld in this pressure boundary l
12 of the casing to vessel.
l 13 The impeller and rotor in the diffuser section are l
( ) 14 serviced from above. They're removed out through the annulus 15 between the core structure and the pressure vessel. The motor 16 is removed from the bottom, removing the cover, the flange 17 connection at the bottom to do that.
18 In order to accomplish that, there are two seals that 19 permit the servicing of the pump. First of all, there is a, 20 what we call, maybe we should say primary seal, first. The 21 primary seal is the fact that the impeller is allowed to lower 22 from the position normally maintained by the thrust bearing by 23 opening a secondary cover on the flange and adjusting both that
- 24 mount that and support that thrust bearing, and allow the back
- 25 seating of the impeller onto the permanent or semi-permanent Heritage Reporting Corporation (202) 628-4888
188 rs 1 part of the pump that is here.
b 2 It closes this gap. It only lowers it a fraction of 3 an inch and closes that gap to form the sort of primary seal.
4 This is rather similar to the back seating of a control rod.
5 MR. EBERSOLE: Is that a hard face seal?
6 MR. KOBSA: I don't know that there is a hard facing 7 on that material right there or not. I don't think it is, but 8 I'm not sure.
9 MR. EBERSOLE: Well, normally the whole motor is 10 carried at full reactor pressure?
11 MR. KOBSA: Oh, yes. This is reactor compressor does 12 communicate down into the pump. It's a wet motor, and reactor 13 pressure and water at reactor pressure is in there.
MR. EBERSOLE: There's normally no contact there at
(]) 14 15 that location.
16 MR. MICHELSON: Now what prevents the motor from 17 being ejected on failure of that well?
, 18 MR. KOBSA: There are hangars that support off of 1
19 lugs on the cover and mount on lugs on the pressure vessel 20 itself in between these pump penetrations.
21 In addition to that, the impeller itself tends to 22 prevent that by the back seating. The back seated propeller 23 went through the thrust bearing to try to hold up the casing if 24 there were a failure of this well, i 25 MR. MICHELSON: There is a trunion of some sort Heritage Reporting Corporation (202) 628-4888
l I that's holding it all up besides.
)
2 MR. KOBSA: Yes. Therc's a pair of rods rather 3 similar to pipe play per strings. About a one and three-4 quarter inch diameter that connect to lugs that are on the 5 pressure vessel in between these penetrations and extend to 6 lugs.'
7 MR. MICHELSON: Have you calculations shown'that the 8 impeller alone would hold the thing together if the well were 9 to fail?
10 MR.-KOBSA: The impeller itself would have no 11 problem. I think that the weak link would be this shaft
- 12 coupling stud. It really goes through a very high stress level i but it would be predicted to, or be able to take loading, but 13
( ) 14 just barely.
15 MR. MICHELSON: In other words, I was wondering if 16 the broad support is redundant to the impeller support?
17 MR. KOBSA: Yes. It really is, although I have no 18 taken credit for that in the design of the external restraint.
19 MR. EBERSOLE: I take it that the stator is canned, 20 isn't it?
21 MR. KOBSA: No.
22 MR. MICHELSON: The stator. Isn't it in a can?
23 MR. KOBSA: No. It's in the water.
24 MR. MICHELSON: The windings are in the water?
25 That's interesting. I thought it would be canned. The stator, Heritage Reporting Corporation (202) 628-4888
I 190 O
(./
1 of-course.
2 MR. KOBSA: The stator, yes. Inside is a screw cage l 3 rotor but the stator, the windings are exposed to the water.
4 MR. EBERSOLE: The cooling water does communicate 5 with the reactor?
6 MR. KOBSA: No. There's a closed cooling water 7 system. The inlet at the bottom. It comes through the pump 8 and out the top into individual heat exchangers and that heat 9 exchanger is mounted so that it will naturally circulate and 10 cool the water in case the pump is stopped. But when the pump 11 is operating there's a built-in impeller that forces that 12 circulation through the heat exchanger during power operation 13 of the pump.
l
( ) 14 MR. MICHELSON: So there's going to be some leakage 15 from the reactor side, though, isn't there?
16 MR. KOBSA: No. This purge water inlet is right 17 here.
18 MR. MICHELSON: No. The shaft leakage.
19 MR. KOBSAs No. There's no shaft leakage.
20 MR. MICHELSON: No. Further up. Where you penetrate 21 the vessel.
22 MR. KOBSA: Where you penetrate the vessel? Well, 23 where we penetrate the vessel along the shaft we introduce this 24 purge flow. The purge flow enters here and comes up through this annulus and past the impeller and out into the flow I ) 25 l
Heritage Reporting Corporation (202) 628-4888 l
191
()
1 stream.
2 MR. EBERSOLE: That's coming from the CRD system.
3 MR. KOBSAt Right.
4 MR. MICHELSON: Now where in the seal between the 5 purge water and the normal cooling water?
6 MR. KOBSA: There isn't any.
7 MR. MICHELSON: Normal cooling water isn't a thousand 8 pounds.
9 'MR. KOBSA: Yes.
10 MR. EBERSOLE: Yes, it is.
11 MR. MICHELSON: Oh, it is.
12 MR. KOBSA: Yes, it 13 It's a reactor pressure.
13 NR. MICHELSON: Reactor building closed cooling water 14 is?
15 MR. KOBSA: No. No. The primary side. This is the 16 primary side of the heat exchanger and this is at reactor 17 pressure. It's on the shell side of the heat exchanger to 18 enhance the ability to naturally circulate.
19 MR. EBERSOLE: It's not quite cool, is it?
20 MR. KOBSAs Pardon?
21 MR. EBERSOLE: It's run rather cool?
22 MR. KOBSA: Yes. At about 150 degrees.
23 MR. EBERSOLE: So, if you lose, it's just a cold 24 water leak? Augmented by a little leakage.
25 MR. KOBSA: Well, no, if you lost the primary side of s
Heritage Reporting Corporation (202) 628-4888
192 1'
(} this then you would start to get reactor water coming on through there.
2 3 MR. EBERSOLE: I know. l 4 MR. KOBSA: But it's part of the reactor pressure i
5 boundary, 6 MR. EBERSOL": There's no quick'way to close that.
7 MR. KOBSA: The primary cooling side of the cooling ,
8 system, not the closed cooling water side.
9 MR. EBERSOLE: How would you stop that leak? !
10 MR. KOBSA: How would you stop that leak?
11 MR. EBERSOLE: Upset this first valve down below?
12 MR. KOBSA: No, you can't. Well, you can do that but 13 only manually from down below. You can't do that there.
()14 MR. EBERSOLE: So, if you have a leak in that high 15 pressure pipe, you just have to de-pressurize the reactor and 16 shut down? ,
17 MR. KOBSA: That's correct.
18 MR. EBERSOLE: You have to shut down?
19 MR. KOBSA: Yes.
20 (Continued on next page.)
21 ,
22 23 24 i
(:)
Heritage Reporting Corporation f (202) 628-4888
I 193 MR. WYLIE You are relying on the cooling water
(}1 2 circuit to keep contamination out of the system if you get a 3 short circuit in that motor.
4 MR. KOBSA: I didn't follow the short circuit part of
. 5 the motor.
6 MR. MICHELSON: Where does the copper go?
7 MR. WYLIE: Where does the residuals go when you melt 8 the winding down?
9 MR. KOBSA: When you blow the winding -- I would 10 think it would tend -- I don't know, to tell you the truth.
I 11 But it would not necessarily migrate into the reactor, but 12 there is nothing to really keep it frora doing this, either. It
( 13 depends upon whether it would build up an excessive pressure or
()14 something like that.
15 MR. EBERSOLE: Is there, in the event we do have --
16 what about 6,000, 7,000 horsepower; 5,000, 4,0007 17 Well, don't worry with it. It is probably about 18 4,000, I bet.
19 Anyway, you can have a heavy explosive short circuit 20 in there. What do you do with the energetic release if you do 21 that? You just intercept it with a breaker?
22 Are you able to account for a heavy short circuit, as 23 Charley says, in that motor without an excessive pressure?
24 MR. MICHELSON: When you blow the pressure boundary 25 with a motor failure, in other words.
I Heritage Reporting Corporation (202) 628-4888 l ;
i 194 !
h1 loop.
MR.-KOBSA: There'is electrical protection in the
.i
. 2
'3 MR. MICHELSON: I know. 'But'even that brings some 4 energy. 'This is not safety grade production unless you -- hava i
5 you put it in safety grade? l 6 MR. KOBSA: No. i 7 MR. MICHELSON: Even though you are protecting the 8 pressure boundary? c 9 MR._KOBSA: As far as that is concerned, that has 10 been examined and the very worst case assumption that we have 11 been able to make on that is a sudden seizing of the pump and
.)
12 that has been factored into the design and it turns out to be a-
! 13 non-limiting condition. The more limited condition being the ;
14 combination of dynamic loads that were designed for and a 15 combination of operational basis earthquake included with the .
16 safety valve transient. Turns out to produce a more limiting ;
i 17 to its criteria load than the faulted conditions including the ;
i 18 seizing of the pump. ,
1 l 19 FR MICHELSON: Have you looked at the uncleared i
20 electrical fault of the motor with the energy source and the j 21 resulting expansion of the water pressurization of the hosing? i 22 Which is the way these faults would go, I would think.
I It would boil the water and pressurize the hosing. l 23 24 Boil the thing out if it was too energetic.
l 25 MR. EBERSOLE: There could be a local detonation.
l
.O a i
i
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i 195 r' l MR. MICHELSON: Yes.
(\/
2 MR. KOBSA: I don't know myself, I have not heard of 3 that particular event being considered at Energy.
4 MR. SAWYER: I don't *hink we have run that specific 5 scenario. We obviously have, from a locus standpoint, non-6 mechanistically looked at all the what-ifs of pressure, of all 7 the pressure boundary failures for the cooling system and the 8 casing itself.
9 MR. MICHELSON: I think what you would say is that, 10 okay, I blow the casing with this fault. The biggest leak I 11 can get is the leakage down through the shaft.
12 MR. KOBSI; Yes, right.
13 MR. SAWYER: That's correct.
( ) 14 MR. MICHELSON: That is on the assumption that the 15 shaft remains intact up in the leakage area.
16 MR. SAWYER: Right.
17 MR. KOBSA: Right. That's about 1000 watts, I guess, 18 of pump power there, and I don't know that that has been 19 evaluated.
20 MR. MICHELSON: Well, hydraulic pressurization effect l 21 woulu probably be the limiting one. You boil the water so fast 22 you pressurize it to several thousand pounds internally because 23 there is no way to relieve it.
24 MR. KOBSA: The only relieving would be through the 25 annulus up into the reactor, I guess, and I believe they have i
Heritage Reporting Corporation (202) 628-4888
196 1 relief on the heat exchangers, also.
{
2 MR. EBERSOLE: Is that a solid water circuit, or do 3 you put any resilience in it to allow for such energy releases?
4 MR. KOBSA: No, it is a solid water.
5 MR. EBERSOLE: Solid water with no air volumes in it.
6 MR. KOBSA: In fact,'that is the intention of it. Is 7 that it be capped off and isolated from the reactor for 8 corrosion protection. That it would dissipate the --
9 MR. EBERSOLE: Well, then, it doesn't have any real 10 room for abrupt energy release from a short. To go anywhere.
11 MR. KOBSA: Other than in the water in that loop 12 itself, that is right.
13 MR. EBERSOLE: So that would stress the loop.
()14 MR. MICHELSON: Why don't we put it on the agenda 15 next time just for some kind of answer.
16 MR. EBERSOLE: By the way, Carl, it would have 17 redundant trips because of ATWS.
l 18 MR. KOBSA: It may be.
19 MR. MICHELSON: What is the source of that cooling L
20 water and injection water?
21 MR. KOBSA: That is the same as the drive -- that is 22 the drive system that provides that cooling water and the perch 23 flow for the drive and for this is the same system.
24 As I guess I started to describe, for the operational 25 maintenance of this pump that the primary seal for the removal Heritage Reporting Corporation (202) 628-4888
197 1 of the internals of the pump -- the motor and the thrush
'\
/~)T 2 bearing from below was the back seating of the impeller onto 3 the seat here.
4 There is a secondary elastomer seal which, in this 5 area, that is inflated upon that operation that will create a 6 secondary seal around the shaft and between the primary seal 7 ano .ne secondary seal be redundant seals to prevent leakage 8 from the reactor during this motor removal.
9 MR. WYLIE: That is at gravity pressure.
10 MR. KOBSA: Yes. You are just at gravity, then you 11 shut down with the reactor head open.
12 Af ter the taotor is removed, then the cover is 13 replaced and the seal is relieved and this is filled with water e
14 and then the impeller and shaft and motor can be removed, if 15 desired, from above.
16 MR. MICHELSON: Is this essentially identical to the 17 Swedish design?
l 18 MR. KOBSA: Yes. As a matter of fact, well, the I
l 19 Swedish design have used a couple different manufacturers, the 20 Swedes have. And the e a couple of manufacturers available for 21 this design, but you are right, it is essentially identical.
l 22 MR. CORK: Irv, if you -- in glancing at the next l 23 four or five charts -- think that you have covered a lot of 24 those points on your presentation, as I glance at it I think i
l 25 you have.
Heritage Reporting Corporation (202) 628-4888
198 1 I would recommeno you go into the operating
(^)T 2 experience.
3 MR. KOBSA: Okay. If there are some questions about 4 those. If you would look over them yourselves, then, I will go 5 on to the slides.
6 MR. MICHELSON: We will revisit the slides on the 7 next meetlag if we find there is so.nething that we wish to ask 8 about. That is all right.
9 MR. KOBSA: Fine. Now there have been, of course, 10 this type of pump -- the wet motor pump -- has been used in 11 circulating pumps in boilers with a lot of experience over the
- 12 years and there have been nearly e hundred pumps in reactors by 1
13 both the Germans and the Swedes in European plants.
()14 And the total experience on those pumps is 15 approaching 600 pump years of operation. In their units, they 16 have somewhere between six and eight pumps per reactor. We 17 have ten, of course.
I 18 There was only one incident of a forced outage due to 19 a pump related problem and that was in the area of a seal l
20 between the pump and what amounts to the shroud support. Where 21 a bolt came out of that and got into the pump and caused it to l
l 22 machine away a part of the fuser section that had to be 1
23 replaced.
24 In the development phase, there have been extensive 25 testing of the pump by the Europeans over the years with l
l Heritage Reporting Corporation (202) 628-4888 l
l l
t
l 199 several pump vendors and in the case in Japan both Hitachi and
(]) 1 2 Toshiba have had independent test loops and testing of their 3 own pumps and then, as you heard here from the gentleman from 4 NRC, the Government -- the MITI -- has funded a facility for 5 testing the pumps in a large, full-size mockup where two pumps
[ 6 can be run in a quadrant of the vessel and it is very similar 7 to our high-flow test facility in San Jose. But it tests two 8 pumps in tandem in that facility.
9 So there has been considerable experience in the way 10 of proving these designs.
11 And unless there are some questions in your mind on 12 the pump or the other things about it in the way of these 13 evaluation sheets, that would conclude my presentation on m
14 Chapter 5.
15 DR. KERR: In your PRA evaluation, what value do you 16 use for the expected frequency of catastrophic vessel failure?
17 MR. KOBSA: The expected frequency of catastrophic 18 vessel failure would be less than one in a million per plant 19 year.
20 MR. SAWYER: I think we are using a number more like 21 less than one in ten million.
22 MR. KOBSA: Well, we may use that, but once it gets l
23 past one in one million then there is no longer a credible 24 design.
25 DR. KERR: Yes, but your goal for 25 REM exposure is l
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200
<~s 1 less than one in a million per year. I would guess a k-)
2 catastrophic vessel failure might lead to 25 REMS of exposure.
3 MR. SAWYER: Right. That is possible, although not 4 automatic.
5 DR. KERR Not automatic.
6 -
MR. SAWYER: Right. We have taken a look at what 7
happened. If you don't look at the thrust loads on the vessel 8 itself, your variance of evaluating needs to know the details 9 of those failure mechanisms on the vessel. But if you do, if 10 you take the containment performance analyses that have been 11 done which do take credit for the limited rate of blow down 12 from the largest pipe break, and you say what if there were 13 basically no limit to the blow down rate. I am going to 1
(~)
(, 14 instantaneously dump the stored energy of the water in the 15 primary system into the containment, it turns out that you will 16 drive containment beyond design pressure but not to two times 17 the design pressure.
18 So it is not automatic. It is not clear that you are 19 going to fail the containment if you do have a catastrophic 20 reactor vessel failure.
l 21 DR. KERR: Some of the experiments at Oak Ridge on 22 vessel catastrophic failure would lead one to believe that the l 23 missiles generated by the process might be significant.
24 MR. SAWYER: There are other considerations, which is 25 very difficult to try to quantify exactly what is going to l
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201 happen there.
{' y 1
. 2 DR. KERR Agreed. And you also, of course, would 3 have an early inventory of fission products for almost 4 immediate availability.
5 So it would seem to me that your vessel failure 6 probability would certainly be -- I assume you are calculating 7 or estimating it to be less than one in a million.
8 MR. SAWYER: Yes. I believe the rationale which we 9 will present the staff is similar to the information which we 10 researched for our GSSAR PRA and the number, as I recall, is 11 something, comes out something like one in 10 million reactor 12 years.
13 MR. MICHELSON: Before we leave the subject of relief
( ) 14 valves, it is my understanding that you do have in your 15 possession now a copy of a memo from Stello to Fraley on the 16 hydrogen detonation and damage to safety relief valves in 17 boiling water reactors?
18 And that also you were sent two or three pages of vu-19 graphs from a presentation that was made to the ACRS on that 20 subject.
21 And I think our only significant question in this 22 regard is, assuring ourselves that you are aware of this 23 particular problem and a short explanation on your part as to 24 why it is a non-problem for the ABWR would be in order.
25 DR. KERR: Carl, I had thought, that we thought at Heritage Reporting Corporation (202) 628-4888 e
~ - _
202 l
() 1 least, this was a problem primarily when one had boron in the v
2 coolant. Is that correct?
3 MR. MICHELSON: No, no. This is the hydrogen 4 detonation that occurred within the safety relief valve in the ;
5 foreign reactor.
6 DR. KERR: Okay. I am confusing two issues, then.
7 MR. CORK: The conclusions in the staff report are e valid for the ABWR in that actuation of the safety relief 9 valver results in a decompression, and therefore you don't get 10 the increased temperatures in the detonation.
11 And that is a true conclusion for the ABWR.
12 MR. MICHELSON: Is that to say then, I guess, that 13 your valve does not have an arrangement that would cause the
(~)s14
(, compression of the gas during the operation and thereby an 15 ignition?
16 MR. CORK: Exactly right. Yes.
17 MR. MICHELSON: Okay. And that is the conclusion I 18 kind of came to in just looking at the outline drawing of it, 19 but I really needed to hear you say it since I couldn't -- you 20 know, there is a lot more to a valve sometimes than a simple 21 outline drawing would show.
22 MR. CORK: All right.
23 MR. MICHELSON: You are aware of the experience?
24 MR. CORK: Yes, we are.
25 MR. MICHELSON: You have looked at it for the ABWR Heritage Reporting Corporation l (202) 628-4888 1
.. a 9
- _ _ - _m - _ ____ _ _ _ _
'203 1 type relief valve.
(]}
-2 'MR. CORK: Yes, we have.
3 MR. MICHELSON: And you be sure yourself that this 4 would not cause a detonation?
5 MR. CORK: That is right. Yes.
6 MR. MICHELSON: Okay. Thank you. '
7 Any other questions on that subject? I think, then,'
8 we are ready to go on to Chapter 6.
9 (Continued on next page) 10 11 12 13 O 14 15 16 17 18 i
19 20 21 l
22 23 l-24
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i 204 1 MR. MICHELSON: By the way, while we're getting tihis
{}.
2 speaker lined up. Something to think about on this valve 3 experience that isn't related directly to the valve. And that 4 is, I had a few conversations with people in the last couple of 5 weeks who had first hand knowledge of this problem, and was led 6 to believe that they're becoming quite concerned about hydrogen 7 accumulation and such things as HPSI and RCSI steam line dead 8 ends and things of that sort.
9 You might want to think about the same problem some 10 more because they are sure, they've convinced themselves now 11 that indeed hydrogen is accumulating and they would rather not 12 have it detonate in some of these dead lakes.
l 13 MR. SAWYER: I'm Craig Sawyer, a Manager of Systems O 14 1nte9retion verzormence sa91 neerine for Genere1 81ectric. Ana 15 I'm going to initiate the discussion of the Chapter 6. And I 16 will then pass the baton over to Mr. Gentry Wade who will 17 specifically address your questions on 6.2, Containment 18 Systems. And then I will pick up the baton again and talk 19 about the emergency core cooling section 6.3.
20 Those two are the two single largest and probably the 21 most interesting sections of this Chapter in any event.
22 (Slide) 23 But to give you an overview of what's contained in 24 the Chapter, itself, there's a section on the safety feature 25 materials that are used. I don't plan to spend any time on it.
Heritage Reporting Corporation l (202) 628-4888
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^T 1 If there are questions that you have about that section, I will
-(J 2 address them or we'll try to answer your questions.
3 These two, as I said, are separate modules which 4 we're going to address in a lot more detail in this available 5 time.
6 In 6.4, there's a discussion of habitability systems 7 which talks primarily about the control room and how you keep 8 the control room available to the operators. At this time, in 9 Section 6.4, there isn't much for you to peruse. All we have 10 done at this time is list the criteria for the design. And in 11 the latest submittal which we're going to provide under Chapter 12 9, Section 9.4.1, we'll discuss the details of the air supply 13 systems and the filtration and so forth for the control room.
() 14 MR. MICHELSON: Does Chapter 9 also describe the 15 chill water systems which you apparently are using ostensibly?
16 MR. SAWYER: Yes. Chapter 9 also covers that. It 17 covers basically all the auxiliary systems.
18 MR. MICHELSON: How about abilities, two ways to 19 think about it. One is people way, but also equipment.
20 Equipment must also have a habitable atmosphere. And that part 21 will be dealt with in Chapter 9, also?
I 22 MR. SAWYER: That's correct.
23 MR. MICHELSON: Thank you 24 MR. SAWYER: Standby gas treatment is 6.5. I'm going l 25 to talk about that as long as I'm standing he)e.
/~T l
\-)
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[:
206
/~'s 1 Then 6.6 talks about in service inspection of class 2 V
2 and 3 components, basically the bottom line is that we're going 3 to be following the rules of the ASME code for in service 4 inspection.
5 And then the last section talks about the nitrogen 6 gas supply system here, the inerting system. There isn't a lot 7 of, there aren't any surprises or new features that are 8 different than inerting systems we've used in the past.
9 So we've chosen for today's discussion basically to 10 concentrate most of all the available time on these two systems 11 with a little bit of information on the stand-by gas treatment 12 system.
13 (Slide)
() 14 The standby gas treatment system. This system as we 15 proposed it for the ABWR is similar in concept to the function 16 of standby gas treatment system that we've used on previous 17 BWRs with a couple of major exceptions.
18 One of these exceptions is that the filter itself is 19 not dual train, but single train. And I have a P&ID which will 20 show that in a moment.
21 And the second thing is the capacity of the standby 22 gas treatment system is reduced and we have taken, the reduced 23 capaci+,y into account when doing our of f-site radiological 24 consequences and in terms of the amount of time it takes to t
- 25 bring the secondary containment be' ' down to a pressure
()
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207
~ 1 less than the outside world.
2 (Slide) 3 P&ID and I'll go back to the main points here. It's 4 a dual divisional system as we have had in the past with the 5 redundant separate power supplies, separate valving, de-6 m'isters, moisture heaters to get rid of extra moisture, and 7 fans. It is single trained in the filter train itself, but 8 redundant exhaust valves powered by separate power supplies to 9 guarantee that we meet single failures from active component 10 design point of view.
11 MR. MICHELSON: What kind of containment pressure can 12 this still be permitted to operate at, since the containment 13 pressure becomes the system pressure after you open the valve?
(} 14 MR. SAWYER: The containment pressure doesn't become 15 the system pressure.
16 MR. MICHELSON: Well, you've got to do some pressure 17 breakdown, and I didn't see a breakdown device.
18 MR. EBERSOLE: Are you going to be talking with or 19 without the pieces you vent through this?
20 MR. MICHELSON: That's what I was sort of getting to.
21 MR. SAWYER: No. This pathway says from the primary 22 containment itself, but that pathway -- I have to get back to 23 you. I do not know the design basis for that.
24 The standard design process is for this to take 25 suction from the secondary containment and handle the leakage O
Heritage Repo: ting Corporation (202) 628-4888
l 208 f}
u 1
2 that you'd get from the primary containment.
MR. MICHELSON: That I follow. I just didn't l
l
,3 understand the primary containment attachment and just what the 4 design conditions were under which it could be used.
5 MR. SAWYER: I do not know the answer to that, Carl. !
l 6 It's a good question. l 7 That's what? Oh, that's for its operation as a 8 normal ventilation system. You're correct. It's not for its 9 operation during an accident.
10 MR. MICHELSON: But it doesn't normally operate, does 11 it?
12 MR. SAWYER: This system is used when you purge the 13 containment.
() 14 MR. MICHELSON: Post-accident?
15 MR. SAWYER: No, normal operation.
16 MR. MICHELSON: So then it's constantly being loaded 17 with contaminants? Atmospheric contaminants?
18 MR. SAWYER: For normal cperation for whatever it 19 finds in the containment.
20 MR. MICHELSON: Whien is dust and all the other --
21 MR. SAWYER: Yes.
~
22 MR. MICHELSON: I don't think it's normally operating 23 during power operation, is it?
24 MR. EBERSOLE: I didn't think you let that system 25 become progressively --
Heritage Reporting Corporation (202) 628-4888
209 I~T 1 MR. MICHELSON: I believe Mr. Wade perhaps has got V
2 the answer. Can you come to a microphone so we can hear, 3 everyone.
4 MR. WADE: We normally exhaust, even whon we vent the 5 containment to de-inert it, we vent through the normal 6 ventilation system, and only when we find high radiation levels 7 of any kind, do we send the gas through the standby gas 8 treatment system. In the normal ventilation of the building, 9 the secondary containment works the same way. The normal 10 exhaust is straight out the stack to atmosphere unless there's 11 high radiation levels, and then you send the material through 1 12 the standby gas treatment before you release it. It's not 13 normally in the circuit. Only when required because of
(') 14 radiation releases.
15 MR. MICHELSON: And never in post-accident from the 16 containment?
17 MR. WADE: Not directly, no.
18 MR. EBERSOLE: Well, now those fans are supposed to l
19 render the secondary containment subatmospheric by what, a 20 quarter inch of water?
21 MR. WADE: Quarter inch of water, yes.
22 MR. EBERSOLE: Now, that's against the thesis that 23 the containment membrane is intact, and one of the more 24 difficult questions is even after an earthquake. Is that 25 correct?
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i
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210 "T 1 MR. WADE: That's correct.
(O 2 MR. EBERSOLE: _And that. involves even such lowly 3 things as cast iron standpipes for drainage, all sorts of 4 potential end leakage problems which you in due course designed 5 to be intact after seiomic interference.
6 MR. SAWYER: 3re you talking about the containment or 7 the reactor?
8 MR. EBERSOLE: The containment membrane structure and 9 the various items that can impact on it. The secondary 10 containment structure. As I recall, there was a lot of trouble 11 with check valves, cast iron stacks, a whole host of things 12 that rendered the secondary containment damageable so you 13 couldn't maintain this quarter inch of water. It's a very l
() 14 troublesome area to make it intact against all these i 15 challenges.
[
16 MR. SAWYER: Of course, you can speak maybe to the 17 design criteria for the secondary containment. But the basis 18 for the system design is a 50 percent in leakage predict.
19 MR. MICHELSON: Well, one thing of course it will not 20 survive is a tornado, and that's admitted in the beginning, 21 because it was a steel structure just like the one I --
22 MR. WADE: No, these aren't sheet metal buildings.
23 MR. MICHELSON: One has to look real hard about how l
24 good is that business of maintaining that quarter inch in the l-l 25 face of external influences and impact damage from post-seismic Heritage Reporting Corporation (202) 628-4888 l
l
211 rs 1 collapse, d 2 MR. SAWYER: The reactor building is designed for 3 seismic conditions.
4 MR. MICHELSON: Yes. In the context of it, it must 5 maintain this leak tightness that gets to be a very complex 6 question. Because there's any number of secondary service 7 systems that will leak.
~
8 MR. SAWYER: Jesse, are you saying that they aren't 9 required to do that, or you just don't think they can?
10 MR. MICHELSON: No, the ground rules say they're 11 required to do it, but the reality of that is questionable.
12 You can't demonstrate it. It can't be demonstrated without 13 just design analysis to see that you have a structure that can 14 withstand all these external influences. It's a hard statement
(])
15 to make to hold that quarter inch of water after such a 16 prodigious thing as an earthquake. I think you almost have to 17 abandon it.
18 MR. SAWYER: 'Do you have anything to add to that, 19 Gentry about the secondary containment?
20 MR. WADE: No. You know it'e a substantial 21 reinforced concrete structure although it's designed for i 22 tornadoes and all kinds of loadings, I can't speak specifically 1
- 23 about the drain system or the cast iron pipes and that sort of 24 thing. But you know, a lot of these drain systems are 25 radioactive and you don't want to be using those kind of joints l
Heritage Reporting Corporation (202) 628-4888
o 212 1 in that system anyway.
2 MR. EBERSOLE: I recall roof drains that came from 3 the roof that went down within the containment as one of those 4 sources.
5 MR. MICHELSON: Well, we'll stipulate that we'll have 6 a tight building.
7 MR. REMICK: Why would you go to a single filter 8 train in the parallel?
9 MR. SAWYER: I think basically the rational for the 10 single filter train was one of simplification and experience in 11 terms of operation. We're adopting three measures for making 12 sure that this filter train is available to do it's duty as a 13 passive device.
() 14 And the kind of problems that we've had in the past 15 have come from not conducting periodic testing effectively.
16 Some of the designs that are out there have a direct water 17 supply which if inadvertently actuated could deluge the
- 18 charcoal. And we solved that problem this time by having 19 basically not a direct connection but a connectable water 20 supply for fire protection purposes.
21 MR. REMICK: Can anybody get access to that water l
l 22 hose? Did the fire you're worried about, I guess, is a fission
(
23 product heating of the charcoal? Is that what it --
24 MR. SAWYER: That heat load the charcoal is designed 25 for. But because it's considered to be a flammable material, Heritage Reporting Corporation l (202) 628-4888
213 1 you have to have a deluge system available. And the complexity 2 of that is that because'it's there, in surveillance and 3 maintenance, it has a habit of being inadvertently actuated, 4 and that's happened in the past. And therefore rendering its 5 usefulness as a device, you know, it basically causes the 6 system to have to go down.
7 DR. KERR: I don't understand these answers to the 8 question that was asked which was why do you use one instead of 9 two. The argument you're using could be used to eliminate both 10 of them, it seems to me.
11 MR. SAWYER: Well, you have to have, you need to have 12 the filter, you need to --
13 DR. KERR: Well, I mean, you need to have two 14 filters. And if the reason you eliminate one is because the
(])
15 deluge system gives you trouble, the reason you have two is to 16 satisfy some sort of reliability criterion, isn't it?
17 MR. SAWYER: If you -- no. What I'm trying to say is 18 that it's our position that it's a passive device, so you only 19 need one. But then you get to the plant technical 20 specifications and if it becomes unavailable because of some 21 surveillance and maintenance difficulty, then you cause the l 22 plant to have to go down. So in the past, we've provided two.
23 So if for operational reasons that if there's a problem in one 24 of them, you can continue operation. That's why I answered the 25 question that way, as opposed to saying, I've got to have two Heritage Reporting Corporation (202) 628-4888 l
214
,3 1 to meet some kind of single f ailure crit.eria.
U 2 DR. KERR: I didn't say single failure, I said single 3 reliability.
4 MR. SAWYER: Yes. It's more reliable if you make 5 more certain of its availability on demand. And by making some 6 design changes, you can do that.
7 -DR. KERR: Don't you have the prerogative of having a 8 grace period within which you can fix it?
9 MR. SAWYER: It's a pretty small grace period, as I 10 remember.
11 DR. KERR: On what basis is it small? Arbitrariness?
12 MR. MICHELSON: Well, that's another whole subject l 13 area of the tech spec.
f'T 14 MR. SAWYER: Yeah, the tecia spec.
Q 15 DR. KERR: But it's critical to the design.
16 MR. SAWYER: I don't remember the time window that 17 has traditionally been used for the unavailability of' charcoal 18 filter, but it isn't much.
l MR. MICHELSON: When we get to the time window, I 19 20 think we'll look for a basis for that time window.
21 MR. SAWYER: As you know, we're attempting to provide 22 some more logical framework for all of our technical 23 specifications, and we're going to be doing that in our tech 24 specs submittal on the AWR that you're going to be seeing next 25 year.
O Heritage Reporting Corporation (202) 628-4888 l
215 1 MR. MICHELSON: Expansion of the window, time window.
2 MR. SAWYER: Yes, expansion of the window.
3 MR. MICHELSON: Are you not now designing that system 4 at least on the theory that you will use it for primary: loop
- 5. venting at some stated pressure?
6 MR. SAWYER: You mean primary containment venting?
7 MR. MICHELSON: Yes.
8 MR. SAWYER: It is not being designed to operate as a 9 har'd pipe system. In other words, if you make it available 10 with the containment at a low pressure, then it can do that 11 function without a problem. But under the conditions that we 12 were discussing this morning, if you wait until the containment i 13 is at design pressure or thereabouts, 10s of psi, let's say, 1
{}
l 14 and then hook this system in, it is not designed for that 15 service.
16 MR. EBERSOLE: Well, if I take the systems that are 17 in being now, and I look at the emergency procedure guidelines, f 18 I don't think-I have systems in place like this that can match l
19 the EPGs. And this is just another one just like that. In 20 short, people are saying they can vent the containment now, but 21 they don't have any equipment with which to do it.
22 MR. SAWYER: I think the rational behind that is if 23 you get so many failures to the point where you find that you 24 need that operation in order to save the containment, then 25 you're willing to risk some damage in this part of the plant as
, ("h
! \/
Heritage Reporting Corporation l (202) 628-4888 1
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216 1 part of what you accept.
)
2 MR. EBERSOLE: I think you're exactly right but 3 that's not very openly said. You're going to blow the guts out 4 in other words, and that'll just be part of the consequence to 5 save the day.
6 MR. SAWYER: Yes.
7 MR. MICHELSON: It depends on where the equipment is 8 located on whether it saves the day or makes the day even 9 worse.
10 MR. EBERSOLE: That's true. It might go to the 11 control room.
12 MR. MICHELSON: Right now, you're not taking any 13 credit for this system to vent containment.
() 14 MR. SAWYER: That's correct.
15 MR. MICHELSON: Now, I heard so many words now I'm a 16 little confused. Why did you think it's perfectly all right to 17 have a single train filter? And then I heard all this good 18 reliability talk and so forth, and --
19 MR. SAWYER: Let me go back and try to give you my 20 perspective on it.
21 From a single failure standpoint, it's use as an 22 engineer safeguard feature, it would have to be-redundant if it 23 were active. It's our position that it is passive.
24 On the other hand, if its reliability isn't high 25 enough, then it makes prudent engineering design sense to have Heritage Reporting Corporation (202) 628-4888
es- 217 b 1 a back up for it, because you don't want to have the plant 2 forced to go down because of its unavailability. And it's 3 those features that we have addressed, in addition to the third 4 one which I didn't get a chance to mention.
5 MR. MICHELSON: If you're going to address the 6 question of reliability, you'll be putting in two trains.
7 MR. SAWYER: That's one way of solvrng that problem.
8 MR. MICHELSON: But you aren't, you're putting in one 9 train. So the inference is I guess it's thought to be so 10 reliable, it's not necessary to use two trains. Is that the 11 case?
12 Why is it one train, which is what you show, passive.
MR. SAWYER: Yes. One train is sufficient. That's
(]) 13 14 our position. And we're trying to simplify the plant and we've 15 gone with the single train as far as the filter is concerned.
16 Now, we have addressed the issues that have come to 17 us through field experience which have caused unreliability in 18 the standby gas treatment system. And by addressing that, 19 we're now comfortable with having this plant go with a single 20 train from an operational standpoint.
21 MR. REMICK: What's the purpose of the spray heaters, 22 space heaters, excuse me, I'm sorry?
23 MR. SAWYER: These space heaters here?
24 MR. REMICK: No, inside the filter train? The X's, I
\l 25 guess.
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218 b<<
1 MR. SAWYER: I don't know the ansker to that.
2 MR. REMICK: Well, my question would be are those ,
3 passive units? Are those passive heaters? Are they needed for 4 the thing to work? Or are they just to dry out the charcoal if 5 it does get moist?
6 MR. SAWYER: I think the answer is that they're there 7 to make sure that at the beginning of any event that you might 8 postulate that the charcoal is always dry.
9 In other words, if those space heaters failed to 10 operate once the event has begun, you don't need them anymore.
11 But I can confirm that and get back to you.
12 MR. EBERSOLE: Space heaters in themselves are a 13 source of combustion problems, aren't they, fire problems?
(])
14 MR. SAWYER: Yes. It's a trade off. You're trying 15 to make sure the charcoal is dry --
16 MR. EBERSOLE: But not burning.
17 MR. SAWYER: -- but not burning. That's correct.
18 MR. EBERSOLE: The stack is a critical aspect of this 19 system. How high is it? What dilution factor are you asking 20 for? What kind of meteorology and so forth?
21 MR. SAWYER: In terms of the plant design or in terms 22 of what we take credit for in the radiological evaluations?
23 MR. EBERSOLE: You put it under a stack to get some 24 benefit out of lifting it. And I don't know what that degree O
\l 25 of benefit is and what you expect -- what are you talking Heritage Reporting Corporation (202) 628-4888
l l
- 219 l O,m 1 ~ about, 200 feet? 50 feet, 600 feet?
2 MR. WADE: The stack is on top of the reactor 3 building.
4 MR. EBERSOLE: So it's just a metal stack?
5 MR. WADE: It's a metal stack but the reactor 6 building is 70 meters tall or 60 meters tall, and the stack is 7 over a 200 foot stack by the time it reaches the --
8 MR. EBERSOLE: Yes, but if it's a short stack, it'll 9 just buckle down behind the tall building.
10 MR. EBERSOLE: It's a rather tall stack.
11 MR. REMICK: To make sure you understand my question 12 about the space heaters, I just want to make sure. I assume 13 they're there to dry out the charcoal in case there's moisture
(])
14 present, but I'm not sure. What I'm really interested in 15 making sure that their presence is consistent with your claim 16 that that's a passive system.
17 MR. SAWYER: And that's what I'll verify for you.
18 MR. REMICK: Okay. Now, the other question I have 19 going back to the host question, I thought one of the concerns 20 you always have with these if you get them ladened with fission 21 products you can get overheating and the potential of fire from 22 fission product heating of the charcoal. And therefore you 23 need some kind of coolability. That can be either air, 24 presumably or I thought water was one of the backup systems.
25 If water is the backup cooling mode in case of a fire Heritage Reporting Corporation (202) 628-4888
1 when the fire starts from radioactive fission product heating 2 how is anybody going to get in there to hook'up that hose?
3 MR. SAWYER: I understand your question, but I will 4 have to get back to you on that.
5 MR. REMICK: Gkay. All right.
- 6. MR. SAWYER: That's all I have on this section. And 7 Gentry is going to talk about the containment section 6.2, and 8 then I'll come back up and talk about the ECS Section 6.3.
9 MR. WADE: We'll start off with just the highlights 10 of Section 6.2. Really, I don't think there's anything in this 11 section that's significantly different than what's been 12 presented in previous analysas.
l Q 13 (Slide) 14 We can look at the containment structure which we 15 went over briefly this morning. Reinforced concrete cylinder 16 to give you some idea of its size, it's 95 feet in diameter, 29 17 meters, and it's about 96 feet high. Completely reinforced 18 concrete cylinder that we've talked about. It's steel lined, 19 steel liner passes up through the diaphragm floor to contain 20 the wall connection across here. Of course, it has a steel l 21 drywell head.
22 And another important feature is the diaphragm floor 23 now is rigidly connected to the containment wall. There's no 24 seal here like there was in some of the Marc II plans. There 25 was a seal there. And the diaphragm floor has a steel liner on l
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221 f3 Q And this portion of the wetted surfaces down in here are 1 it.
2 all stainless steel, 3 And we have the upper drywell suppression chamber air 4 space, suppression pool that we've tal.ked about before, and the 5 lower drywell. And I believe we did discuss the venting 6 system this morning.
7 One thing I want to point out is that the containment 8 liner does pass through the bottom of the pedestal. And I told 9 you there was 1.6 meters of concrete cover over here. This is 10 the containment liner that continues on across to the other 11 side. So it's flat on the bottom.
12 (Slide) 13 The containment is structurally integrated with the
(])
14 reactor building. You are familiar with integration on the top 15 slab, the mark III containment had the refueling pool girders 16 integrated with the top slab. This goes a step further and 17 integrates the reactor building floors with the containment l
l 18 structure at three levels, plus of course, there is a common 19 base mat.
20 The design loads are similar to what we've used
! 21 before. Dead and thermal load, Loca dynamic loads, te talked 22 about the design of the vent system this morning being similar 23 to what we have in Marc III. There have been a question of CO 24 and chugging loads of what would happen with more back 25 pressure. The Marc III system is essentially zero back
! 3eritage Reporting Corporation (202) 628-4888 )
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~ 222 1 preseure. The ABWR with a design pressure of 45 psig could 2 have higher design pressures.
3 We did run a special test for Tokyo Electric Power 4 Company in the Marc III test facility at San Jose. And we've 5 determined that the trend for both chugging and CO loads is 6 down with increasing back pressure. So the dynamic CO and 7 chugging loads for the ABWR are less or will be less than they 8 were for Marc III, although we don't necessarily take 9 significant credit for that in the design.
10 The SRV loads, we use as a basis the quencher loads 11 that we've had before from our other full scale tests and tests 12 at operating plants.
13 Pool swell configuration is similar to the Marc III
(])
14 in that you have horizontal. vents discharging on one side of a 15 pool, although this pool is a little wider than the Marc III 16 pool so you have the growth Of the air bubble and the growth of 1? the pool swell in this region very similar to what we had in 18 Marc III.
19 The design pressure set at 45 and (-) 2. The (-) 2 20 is set primarily by keeping the liner attached to the concrete 21 through the anchorage system. The liner itself has no 22 pressure, is not depended upon for pressure carrying 23 capability. This is all taken by the reinforced concrete 24 structure for positive internal pressures. For negative G
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1 attachment to the wall.
2- MR. MICHELSON: Now, it's my understanding you have 3 no vacuum relief 'o the secondary containment?
4 MR. WADE: That's true. It's not required.
5 MR. MICHELSON: Now, this leads to a couple of 6 questions.
7 When doing a routine shutdown with this type of 8 plant, at what point do you de-inert and start ventilating the 9 containment? What are the reactor conditions when you start 10 de-inerting and ventilating the containment?
11 MR. WADE: I can't give you the operating numbers for 12 when we do that.
() 13 MR. MICHELSON: Well, let me give you a problem that 14 we'll discuss then the next time we meet.
15 The problem is that if there are any openings in this 16 containment at a time when you could get a loss of coolant 17 accident, a severe one, you will release a portion of your non-18 condensable gas in the process of blowing it out through the 19 ventilation ducts while you're trying to get the ducts closed 20 and so forth.
21 So you have to look very carefully at how many non-22 condensables you lose because now the negative pressure you can 23 generate with a water spray is quite a bit more. And so it 24 depends upon at what point you think you're going to de-inert kJ 25 and start ventilating the containment while the reactor could l
I Heritage Reporting Corporation (202) 628-4888 l
-s 224
~
1 still have a fairly significant accident.
2 MR. SAWYER: As I remember, we're not planning 3 anything revolutionary in this area. And as I remember, 4 typically for inerted plants, you can start de-inerting when 5 the plant is at 25 percent power and you have 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to do 6 it.
7 MR. MICMELSON: Yes. But our present designs, do 8 they not have vacuum relief?
9 MR. SAWYER: The Marc I has vacuum relief mainly 10 because you're worried about potential collapse of the torques.
11 That's a different scenario.
12 MR. MICHELSON: How about Marc II. Does it have
{) 13 vacuum relief? I didn't realize, I thought they all had vacuum 14 relief.
15 MR. WADE: No, I don't think the Marc II freestanding 16 steel might need it, but I don't think the lined concrete ones 17 have it, need it.
18 MR. MICHELSON: How about Marc III?
19 MR. WADE: Marc III's have a vacuum relief. They're 20 at a much lower pressure.
21 MR. MICHELSON: Maybe the problem belongs on Marc IIs 22 and IIIs too, but I'm going to ask it for ABWR. I'm going ask 23 if you open the ventilation system up and you experience a loss
- 4 of coolant accident, are you assured you can hold enough non-
! () 25 condensable gas so that when you spray the containment, you l Heritage Reporting Corporation (202) 628-4888 L
225 1 don't draw more than the design negative pressure?
2 It wasn't clear from reading the document that you 3 took into account loss of non-condensables in the process of 4 the~ accident. I think you do start ventilating long before 5 your down in power temperature and pressure very far. It 6 appears there's several hours in which you would be vulnerable 7 to that kind of an accident.
8 MR. WADE: And we are accounting for uninerted 9 containment in our PRA, and for this window. I just don't 10 remember, as I said, the exact details of what the proposal is 11 or at what power level you can start de-inerting.
12 MR. MICHELSON: The other question that's going to
{} 13 14 come up eventually on your containment isolation.
propose to open the containment while you still could have a If you 15 significant loss of coolant accident, then you have to assure 16 us tnat you could close the containment isolation valves again.
17 Besides the loss of non-condensables, can you even get the i
18 valves closed. And what kind of spec are you writing and so 19 forth to, what kind of interface requirement are you putting on 20 it, or what are you doing to assure that.
21 MR. EBERSOLE: T'is n matter of mounting the liner to 22 the concrete, inevitably raises the* question of the thermal 23 experience when you dump a lot of hot water on the steel and 24 produced a buckle. Because the concrete is not hot. I take it
() 25 you accon.modate that by some allowance in the anchors and Heritage Reporting Corporation I
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1 offsets the steel to permit that buckling to take place.
2 MR. WADE: We have calculated the load on the steel-3 due to it getting warmer for things like that, and I think it 4 remains within the limits of design. There are no special 5 pleats or anything in the design to accommodate that. It's not 6 required by design.
7 MR. EBERSOLE: What you don't want to do is pull out 8 the pins.
9 MR. WADE: You know, the containment really doesn't, 10 although it's inerted, it doesn't really operate at a very high 11 pressure. It's a fraction of a psi. And I would say that you 12 know when we're de-inerting, we're pumping air in as fast as,
(} 13 and trying to get three or four volume changes of air through 14 the building to sweep out the nitrogen. So we can evaluate 15 that particular point.
16 MR. MICHELSON: What is the normal containment 17 operating pressure during full power operation? What pressure l
l 18 are you trying to hold in the containment?
19 MR. WADE: We do the analysis at 15 and a quarter or 20 a half psi.
21 MR. MICHELSON: You mean you're keeping a positive 22 pressure.
23 MR. WADE: Positive pressure to make sure that any J nitrogen leaks outward instead of inward,
{ 25 MR. MICHELSON: About a pound or so?
! Heritage Reporting Corporation (202) 628-4888
l
- 227-I 1 MR. WADE: I think that's a number we used for the 2 containment analysis.which is a little higher than the-3 operating point'that they used_to make sure that we have the 4 maximum non-condensables in the containment to calculate the 5 maximum pressure.
6 MR. EBERSOLE: Is the water volume, what's the~-- is 7 that carbon steel?
8 MR. WADE: Water volume's all stainless steel.
9 MR. EBERSOLE: Is it? Okay.
10 MR. WADE: It's all stainless steel, even in the vent 11 system, anything in the wetted surface that's down in this part 12 o'f the pool is stainless steel.
() 13 MR. EBERSOLE: Okay, great.
Not because it's less expensive, but 14 MR. WADE:
15 because it eliminates coating problems and getting rid of 16 coatings and clean-up and sprays and all sorts of long term 17 operating problems.
18 This just says that the area,-the volume of the upper 19 and lower drywell and the suppression chamber have a, there's a 20 ratio of this to that of about .8. These numbers are similar 21 to what we've had in the past on the Marc I and II containments 22 as far as the volume ratio goes.
23 MR. MICHELSON: Are there sprays in the suppression 24 chamber?
25 MR. WADE: Yes, there are. And the drywell, and the Heritage Reporting Corporation (202) 628-4888
() 1 upper drywell.
2 MR. MICHELSON: With the presence of these sprays, 3 you've looked at the decompression _ rates when you hit the thing 4 with full steam?
5 MR. WADE: Yes, we have. That's one of the accidents 6 we've analyzed as far as the vacuum breaker, which I'll get to 7 in just a minute.
8 I think we've already covered everything that's on 9 this slide including the additional full scale testing program 10 that we completed for Tepco a year and a half ago.
11 (Slide) l 12 Now, the calculated peak pressures in the drywell and 13 the wetwell and the temperatures are shown on this chart. Our
(])
14 peak drywell pressure occurs in the short term which is 15 indicated by the 2. And it occurs for a feedwater line break.
16 We find that with the ABWR design putting the flow 17 restricters in the reactor pressure vessel nozzles, the 18 feedwater break and the steamline break are very similar.
19 Sometimes one controls and sometimes the other.
20 But this gives us our 15 percent margin as required 21 in the design rules. We have slight change on this one slide.
22 That's supposed to be an, 8. The peak calculated pressure is 23 338 degrees. The temperature, I mean, is 338 degrees for this 24 accident. This happens to occur in the steamline accident in 25 the drywell in the short term. The very short duration peak, Heritage Reporting Corporation (202) 628-4888
229 1 certainly the steel has been analyzed for this pressure, this
.2 temperature but the peak is over so fast, it really has no 3 affect at all on the concrete structure itself.
4 So this does put the steel in compression as it 5 expands.
6 MR. MICHELSON: These are guillotine breaks, I 7 assume?
8 MR. WADE: Yes, they are. Guillotine breaks in 9 .either the feedwater or the steam line. There's extensive 10 discussion in the document about the flow rates out the breaks 11 and what the break areas are and that sort of thing.
12 I didn't intend to go into those in detail. But they
(} 13 are similar to uhe analysis we've used before. The models 14 we're using, M3, Citgo 5 and for the short term, and Super Hex 15 04 for the long term are models that we used in the Marc III 16 licensing process. There may be updated versions of them but 17 they're the same basic models. And have been accepted for 18 licensing.
19 The design temperature on the wetwell is 219 degrees 20 Fahrenheit in the pool, and our analyses show that we reach 21 temperatures in the 207 degree F. range. This provides a 22 margin of 5 degrees to saturation at no back pressure.
23 This is a fairly straightforward compared to what 24 we've done before.
25 (Slide)
Heritage Reporting Corporation (202) 628-4888 9
() 1 Now, this morning, I talked about the vacuum breakers 2 which are located as you can see in this picture up at this 3 level, they go between the suppression chamber air space and 4 the lower drywell. These penetrations are separate from the 5 vent system. They have nothing to do with it. This is a 6 designed feature that was added of course to the Marc II 7 containments.
8 And there are eight 20-inch vacuum breakers that make 9 this connection. The calculations of pool swell and everything 10 indicate that the vacuum breakers remain closed during pool 11 swell. They also remain closed during chugging, which is not a 12 real surprise since they are on separate penetrations and don't 13 see the pressure fluctuations that might occur in the vent
(]}
14 system here itself.
15 About the only time they open is on spray actuation 16 either an inadvertent actuation of the spray in the drywell or 17 the wetwell. There could be an actuation of the spray in the 18 wetwell if the SRVs were discharging for a long period of time 19 through the suppression chamber and you wanted to mix and cool 20 the air above that. Or of course flow out the break.
l 21 MR. MICHELSON: They are the normal testable variety 22 like you've used in the past?
23 MR. WADE: Yes. They are the normal testable variety 24 with lights in the control room so you can tell if they're open 25 or closed. You can test them monthly.
Heritage Reporting Corporation i (202) 628-4888 l
-- 231 b 1 MR. MICHELSON: We've had a great deal of trouble 2 with those valves because of the bad atmosphere inside the 3 wetwell. Are-these going to have their solenoids and so forth 4 all inside the wetwell?
5 MR. WADE: At the present, the valve is all located 6 in the wetwell airspace, yest
,7 MR. EBERSOLE: Are you going to monitor this 8 containment by pressure differential which is maintained and 9 observed as an evidence of non-excess leakage? Do you know 10 what I mean?
11 MR. WADE: Well, since it always operates at a slight 12 positive pressure, you can determine if there is leakage from
(} 13 the containment by that, by monitoring the --
14 MR. EBERSOLE: Yes, that's the whole thing.
15 MR. WADE: Yes, that's the whole thing.
16 MR. EBERSOLE: What about between the various 17 compartments of it?
18 MR. WADE: There is no, there is no plan now to 19 monitor any difference in pressure between -- well, they're 20 can't be any between the upper and lower drywell, or between 21 the drywell and the suppression chamber airspace.
22 MR. EBERSOLE: Don't you have continuous control air 23 end leakage in the apparatus that's inside that you need to get 24 rid of or else recycle the flow path?
25 MR. WADE: I haven't looked at all of the control Heritage Reporting Corporation (202) 628-4888
_,~
232 l
<-) 1 systems. I don't believe there's much control air used in the 2 drywell.
3 MR. EBERSOLE: I recall at the Brown's Ferry Plant, 4 it was decided eventually to pump the air from controls out of 5 the containment from whence it went back to avoid that problem 6 because they had a constant build-up.
7 MR. MICHELSON: You also get a nitrogen dilution if 8 you --
9 MR. WADE: Well, that's one of the things that we 10 have to do of course is use nitrogen for any of the systems 11 that are putting air or putting gas into the containment on a 12 routine basis.
13 . MR. MICHELSON: I thought the relief 'ralves were air
{~)
14 operated, though, with a nitrogen backup. Did I misunderstand?
15 Oh, they're nitrogen operated.
16 MR. WADE: Yes.
17 MR. MICHELSON: Okay. No external nitrogen, or the 18 nitrogen comes from a bottle outside the containment, I guess.
19 MR. WADE: Yes.
20 MR. MICHELSON: But no air into the containment for 21 any reason, no air operated valves or anything?
22 MR. SAWYER: Nope. It's a misnomer. It's called 23 instrument air system but it really should be instrument 24 nitrogen system.
() 25 MR. MICHELSON: Yes. As a matter of fact, the P&ID Heritage Reporting Corporation (202) 628-4888 i
233 O 1 did show it, yes. Okay. We'l) fix that some time.
2 MR. WADE: And we have examined the many types of 3 accidents that can go on in the drywell or wetwell which might 4 cause depressurization. And the maximum that we have after a 5 LOCA, a feedwater line break, is (-) 1.5 psi, which is well 6 below the design pressure. And the vacuum breakers between the 7 wetwell and drywell are sized for that.
8 Now, for those accidents like the SRVs discharging to 9 the pool, if you turn on, inadvertently turn on the spray 10 system or do it on purpose, the maximum calculated differential 11 pressure between either of the containment compartments, the I
12 suppression chamber airspace or the upper drywell and the l
{} 13 outside at.mosphere is 1.8 psi. And since this is lower than the design pressure of 2 psi, there is no need to provide a 14 15 vacuum oreaker between the reactor building and the interior of 16 the containment.
17 MR. MICHELSON: But your upper part is still designed 18 for 2 pounds negative differential?
19 MR. WADE: All of the lining is designed for 2 pounds 20 negative.
21 MR. MICHELSON: And in the lower part, you have a 22 maximum of 1.5 but in the upper part, you've got a maximum of 23 1.8. Is that correct?
24 MR. WADE: No. Well, no. Well, yes, the maximum
() 25 occurs across 1.8 here. There is also a differential --
i i
l Heritage Reporting Corporation (202) 628-4888 L
234 7-MR. MICHELSON: So I'm looking at a 2 pound design 1
2 and a 1.8 calculator?
3 MR. WADE: Yes.
4 MR. MICHELSON: 'Which is not what I'd call a 5 comfortable margin anymore.
6 MR. EBERSOLE: I wouldn't, either. What are the 7 margins of conservatism against implosion failure? At what 8 pressure would it implode?
9 MR. WADE: When would the liner come off?
10 MR. EBERSOLE: Yes. Oh, it's bonded to the concrete.
11 MR. WADE: Yes. It'd come off at 4 or 6, you know.
l l 12 MR. EBERSOLE: Yes, it's not prestanding steel. It's
{) 13 14 locked to the concrete.
MR. WADE: Yes. The concrete will take the full 15 pressure. Concrete is not --
16 MR. EBERSOLE: Sure. On a free standing steel can, 17 it's different.
18 MR WADE: Yes. You'd have buckling considerations.
19 But here because we have so many anchors, buckling is not a 20 consideration.
21 MR. EBERSOLE: Right.
22 MR. MICHELSON: This disadvantageous geometry up 23 around that upper head, I'm not sure that your answer is true,
(
i l 24 Buckling would be a serious consideration up there, inverse
() 25 curvature up there, a well made liner we're talking about now.
Heritage Reporting Corporation (202) 628-4888
235 O
\- '
1 MR. WADE: Right.
2 MR. MICHELSON: That's all steel though, up there, 3 isn't it?
4 MR. WADE: Yes. This is steel --
5 MR. MICHELSON: The lin'ir starts only in the box 6 part?
7 MR.. WADE: Well, the liner starts right here where 8 the flange joins the drywell head to the extensiis f Of the 9 drywell to the liner really starts there and goes around.
10 MR. EBERSOLE: So it applies just to the dome.
11 MR. WADE: Yes.
12 MR. MICHELSON: The dome is all steel.
13 MR. WADE: Yes, but that's an inch and a half or can 14 be inch and three quarters steel. And I think it can be 15 designed to resist. It's about 33 feet in diameter, a dished 16 head, ellipsoidal head, it can be designed for that.
17 Bypass leakage, there's usually a question about 18 bypass leakage between the upper drywell and the lower drywell.
19 We've mentioned before the rigid connnection between the 20 diaphragm floor and the containment and the steel lining and 21 the reactor pressure vessel pedestal itself is all fabricated 22 steel. Not only the inner and outer shells but the vent system 23 down through it is a steel shell, and of course it's filled 24 with concrete later on.
l () 25 So we essentially have a steel interface between the l Heritage Reporting Corporation (202) 628-4888 I ._ . . , _ _ _ _ _ _ _ . _ . - - _ _ _ . _ _ _ . . _ _ _ _ _ . _ _ . _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _
236 1 drywell and the suppression chamber. And we don't look to that 2 leaking very much, if at all. But the suppression chamber 3 sprays operate off two of the RHR systems. And the spray 4 bypass capability has an A over root K of over .05 square feet 5 which is the required value that was established for Marc II 6 containment which has possibly a similar configuration.
7 So it meets the requirements as stated in the 8 Standard Review Plan.
9 MR. EDERSOLE: Why didn't you put those discharge SRV 10 pipes down the outer containment wall, imbed them in concrete 11 like you did the main pressure reliefs?
12 MR. WADE: Well, there's something to be said for 13 being able to get to pipes and inspect them and that sort of
({}
14 thing.
15 MR. EBERSOLE: I agree. True.
16 MR. WADE: And all, and --
17 MR. EBERSOLE: But you have to not use that argument 18 for the main steam release -- I mean the main bypass release.
19 MR. WADE: There's, you know, it didn't show up on 20 the drawings but we have access to the upper drywell. There's 21 a two meter diameter hatch and there's inspection platforms in 22 here not only to work on the valves, but to provide very good 23 access to all the safety relief valve discharge lines. And 24 they are built to very high quality but they're also very 25 serviceable, very accessible for inspection.
Heritage Reporting Corporation (202) 628-4888 i
_ _ - - - I
237 O 1 MR. EBERSOLE: Are they carbon steel?
2 MR. WAD 3: Yes, sir.
3 MR. EBERSOLE: They don't rust.
4, MR. WADE: Well, parts under the water I think are 5 stainless steel.
6 MR. MICHELSON: The environment'there is very 7 nonconductive protecting the piping. That's a human hot 8 environment Et all times in the wetwell. Experience on other 9 PWRs has been that that's an area that unless you're awful 10 careful with coating the pipe and so forth, it's going to rust 11 away.
12 DR. KERR: Well, does it have any oxygen?
13 MR. MICHELSON: Sure. It's got air. No, they're
)
14 inerting, that's true. The old experience with inerting, it 15 should help. You're right.
16 DR. KERR: Yes, that should help.
17 MR. MICHELSON: Yes, it should help a lot.
l l 18 MR. WADE: We'll go back to the colored slide over 19 here which you saw this morning with a brief discussion of a 20 secondary containment. It completely surrounds the primary 21 containment, as you can see from this. cutaway view with the 22 primary containment being the dark area.
23 There have during refueling operations when the 24 drywell head is off, the secondary containment does serve as I) 25 the containment against release of fission products. Now, we Heritage Reporting Corporation (202) 628-4888 l
p 238 O 1 don't take the drywell head off and open the hatches into the 2 containment until the reactor vessel itself is depressurized.
3 But the secondary containment, we get into it with airlocks.
4 It has sealed penetrations or electrical penetri -ns .
5 Fluid systems that go to the clean area have 6 isolation valves, or if they're drain lines, they might have 7 water seals. We're trying to maintain a quarter inch of water 8 pressure in there and have the leakage be less than fifty 9 percent of the secondary containment volume per day. So it 10 doesn't take much of a water seal to do that.
11 [ Slide) 12 It operates at a negative pressure relative to the 13 primary containment and the reactor building clean zones. Now,
(])
14 the document'contains pictures of the reactor building clean 15 zones. And I didn't want to go through all those floor by 16 floor but I can show you this and give you a copy. Well, I 17 guess it's already in the document.
18 There are clean zones within the reactor building for (
19 electrical equipment areas, diesel generators, reached back 20 areas and that sort of thing. And so this secondary l
21 containment which is inside this dotted line and inside this 22 hatched area all operates at a negative pressure relative to 23 either that space or that space. So flow is to the secondary 24 containment and then out through the exhaust system.
25 MR. EBERSOLE: Tell me, can't you reasonably argue Heritage Reporting Corporation (202) 628-4888 t
l
73 239 V 1 that you don't need integrity of secondary containment in the 2 presence of earthquakes and tornadoes because you don't have a 3 coincident accident? Are you going to do that? It would save 4 a lot of trouble.
5 MR. WADE: It would save a lot of trouble, yes. If 6 we could show that that's a doubling up of accidents.
7 MR. EBERSOLE: Well, I know that the older practice 8 was to argue that in certain case of tornadoes, you didn't need 9 secondary containment because you're not going to have it.
10 It's going to be gone.
11 MR. WADE: Yes. But some of the old secondary i
12 containments were steel frame buildings with siding on them.
{) 13 This is a completely reinforced concrete structure.
14 MR. EBERSOLE: That's true. Yes, but the question 15 is, do you need to invoke those arguments and cope with the 16 problems that go with it.
17 MR. WADE: It would be very helpful to counteract 18 some of the arguments about leakage of drainage systems and 19 some of the other --
l 20 MR. EBERSOLE: Well, it's a standing argument, you 21 better not have a local and an earthquake, and you better not 22 have it with a tornado. Se it follows the reason I believe 23 that you don't need these capabilities, for the secondary l
24 containment.
( 25 . MR. SAWYER: For those specific external events.
i l
Heritage Reporting Corporation (202) 628-4888 l
240
()
l 1 MR. EBERSOLE: Yes, right. You just invoke that 2 they're not coincident. No way.
3 MR. WADE: The thing that we have done and has been 4 done in all the previous plans, of course the ECCS compartments 5 do have high energy lines in them. We analyze for breaks in 6 these compartments. The compartments are vented through the 7 various paths to the steam tunnel and then out to the turbine 8 building.
9 And this is a case where the analyses show that this 10 can be done as a routine basis without -- well, not on a 11 routine basis. Excuse me. Which can be done without exceeding 1
12 the radiation dose limits.
{} 13 MR. EBERSOLE: You lose the secondary containment in the event of a turbine accident, don't you? Or do you?
14 15 MR. WADE: The turbine building to my way of thinking 16 doesn't really have a secondary containment.
17 MR. EBERSOLE: Oh, I know that. But does it 18 interface with the secondary containment to the extent it's 19 susceptible to damage.
20 MR. WADE: We have to demonstrate that failure of the i
21 turbine building will not impair the functional operability of 22 our seismic category one structure. We have to demonstrate 23 that as part of the design process.
24 MR. MICHELSON: Well, this orientation I guess is the
() 25 same one you're going to use, or perhaps not.
1 Heritage Reporting Corporation (202) 628-4888 l
l
241 1 MR. WADE: In line, --
2 MR. MICHELSON: This one here shows that the peril of I
3 a building turbine, but I guess maybe you're not necessarily 4 doing that.
5 MR. WADE: I think that we are studying the in-liner 6 arrangement which is I might say the preferred orientation in 7 the United States and the one that is required by the EPRI 8 requirements document.
9 The Japanese have used both kinds but sometimes their 10 sites are rather small, and by using this turbine building 11 arrangement, they can get a little on a little smaller site.
12 MR. MICHELSON: Does that mean that there might be
(}. 13 two standard plants, one with a tangential turbine and one in 14 line? .
15 MR. WADE: I don't believe that there will be~two 16 standard plants as such. The primary interface between the 17 reactor building and the turbine building is the steam tunnel 18 and it remains in the same place, same orientation, same-19 elevation. So it makes no difference on the reactor building 20 that we're studying today.
21 MR. MICHELSON: You mentioned that you're providing 22 vending from the high energy lines to I think you said the 23 steam and feedwater chase. In the case of the RCSI turbine, at 24 least in the way I'm looking at here, it seems to be very far 25 removed from the steam chase. How do you vent that room?
Heritage Reporting Corporation (202) 628-4888
4 1 MR. WADE: I think that that picture's very 2 misleading. The RCSI turbine is up in the zero 90 degree 3 quadrant of the reactor bui'1 ding and it is very accessible to 4 the steam tunnel.
5 MR. MICHELSON: This may not even be a good drawing 6 we've been looking at then, because in this one, it's on the 7 basement elevation right next to the RHR.
8 MR. WADE: That's true, but those --
9 MR. MICHELSON: And the steam tunnel's at right 10 angles to it 90 degrees.
11 MR. WADE: Yes, but those, remember those things on 12 the basemat, all those ECCS rooms, the RCSI turbine and all 13 that, have vertical pipe chasers that come most of the way up
(])
14 the side of the containment so the pipes can enter the 15 containment at about this level. And the RCSI turbine you're 16 talking about happens to be in this cutout section and in fact, 17 it's very easy to get a connection from the steam tunnel over 10 ' to these pipe chasers.
19 MR. MICHELSON: This is a different drawing, then, 20 because this one is not in the first 90 degrees.
21 MR. WADE: The artist has a little trouble. And I 22 might say, someday if you look at the pump for the condenser
< 23 cooling water on that drawing and take a look at it real 24 closely, you'll see it's in a very strange position.
- 25 MR. EBERSOLE
- I don't think we have a copy of that Heritage Reporting Corporation (202) 628-4888 l
243 O'
1 color picture, do we?
2 MR. MICHELSON: No. No, we asked for copies of it,
- 3 and we will get the right stuff to work with.
4 MR. REMICK: Looking at your last bullet there and 5 recalling our previous conversation, and looking at what you've 6 done in the secondary containment to make that a much more 7 substantial building over some of the previous designs, I 8 realize that one can save money by putting in one standby gas 9 treatment train, but it seems like it's insignificant when you 10 realize that that's your last protection of the public in case 11 of release out of the primary and the secondary.
12 I guess I just can't see what's driving you toward
{} 13 14 one train, even if it's nothing more than to pacify somebody.
It seems like a trivial change from a cost standpoint, but a 15 significant change in protection, perhaps.
16 MR. WADE: Well, it's an investment protection 17 decision and it's not only the cost of the train, but it takes 18 up a lot of valuable building real estate. You have to build a 19 bigger building to get that stuff in.
20 MR. REMICK: But you have two de-misters and all that 21 there. All we're talking about is what, something three feet 22 or four feet by four feet by 15 feet long, something like that?
23 MR. WADE: These are seven or ten meters long and we 24 have to have a space that's four to five meters wide. Yes, it
() 25 becomes a fair volume to enclose that.
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244 O 1 MR. EBERSOLE: .The Japanese were very proud in this 2 connection.about not having chinchey little buildings wherein 3 maintenance was difficult. You seem_to have compacted this 4 plant and taken pride in it, and perhaps to the degree you 5 challenge the thesis of having large spaces would make 6 maintenance easy. -
7 Is that impression wrong? They took_ great pride in 8 the great open buildings that they had from the standpoint of 9 maintenance?
~
10 MR. WADE: Yes. And we take great pride in the 11 maintainability of this building too.
12 MR. EBERSOLE: Yes, but you also claim, you know, you 13 save space.
[}_
14 MR. WADE: Well, for instance, down here on the 15 bottom floor, immediately when you come out of the equipment 16 hatch, immediately adjacent to that is a whole room devoted to 17 control rod drive maintenance, both the electric part and the 18 hydraulic part.
19 MR. EBERSOLE: I don't need the details. I'm just 20 talking about kind of a policy.
21 MR. WADE: They have looked very closely at that.
22 And this whole next-to-the-bottom floor over all the ECCS rooms 23 is mostly open so that they can get down in there and pull up 24 the equipment and work on it.
( 25 The upper drywell, I don't know if any of you have Heritage Reporting Corporation (202) 628-4888
245
~
1 ever been in Brunsbuttel, but some of these German plants when 2 you walk into them, it's like walking into this room. There's 3 a reactor over there and a few pipes, but it's not like walking 4 into a submarine machinery room.
5 You get the same thing here where you can walk in, 6 there's a big open area, there are monorails to life all the r
7 SRVs out to work on these things. The Japanese in Tepco, they 8 have looked these over extensively and they feel that it's a 9 very maintainable plant.
10 MR. EBERSOLE: What about the fuel handling business 11 and the cast drawbacks, and etcetera.
12 MR. WADE: Well, I'll tell you, it's more than 30 13 feet from grade level up here and over and back down into the
)
14 --
15 MR. EBERSOLE: And you accommodate the cast drop 16 problem?
17 MR. WADE: We can accommodate the cast drop problem 18 by using the double rig safe cranes that we've used in the 19 past.
20 MR. EBERSOLE: Is that your intent?
21 MR. MADE: Yes.
22 MR. EBERSOLE: In other words, you're not going to 23 drop it?
24 MR. WADE: Yes.
() 25 MR. EBERSOLE: Yes, but you do it, though, in some Heritage Reporting Corporation (202) 628-4888
'1 way which is kind of left handed. For instance, if you' drop 2 the cast in the pool, will it bust the bottom out?
3 MR. WADE: No. The cast never comes over the main 4 pool. The cast comes over a separate cast loading pool.
5 MR. EBERSOLE: Okay. So it never gets there.
6 MR. WADE: Yes.
7 MR. EBERSOLE: It can only fall in the high vertical 8 chase?
9 MR. WADE: True. And we try and put load limits on 10 the cranes so you just can't carry the cast indiscriminately 11 any place in the upper refueling area. This refueling area is 12 very similar in its operation to the Marc I and Marc II type 13 plants with the proviso that there is a separate cast loading
({} ,
14 pool in here.
15 Any other questions? If not, it's up to the Chairman 16 whether we break or Craig talks.
17 MR. MICHELSON: Well, I guess it's probably a good 18 time for a break since it's quarter to 4:00. We'll take a 15-19 minute break until 4:00 o' clock.
20 (Whereupon, a brief recess was taken.)
21 MR. MICHELSON: I think we'll readjourn and proceed.
22 You're going to do the emergency core cooling systems, next, 23 sir?
24 MR. SAWYER: Yes, and then Chapter 15 after that.
l 25 MR. MICHELSON: Okay.
Heritage Reporting Corporation (202) 628-4888
247 0 1 MR. SAWYER: During the break, someOne absconded with 2 a package I had sitting here.
3 MR. MICHELSON: That's nice.
4 MR. SAWYER: There was a package I had here of my 5' view graphs for the Chapter 6.3.
6 (Discussion regarding missing view graphs off the 7 record.)
8 MR. SAWYER: Okay, 6.3. I thought I'd start off this 9 section 6.3 to orient everybody. This is a chart I've given 10 before but just in case there are some who haven't seen it, 11 this just gives you some background information on typical l 12 BWR\4 ECCS networks, typical view of BWR\5 and BWR\6 ECCS
(} 13 networks, and what we are proposing for the ABWR.
14 And it compares it both in terms of number of divisions, 15 the BWR\4s are two, BWR\5 and \6 are three, but in one of the 16 divisions, there's only one component and that's the high 17 pressure core sprays. And the ABWR there are three complete 18 divisions, complete in a sense that there is high pressure and 19 low pressure ECCS function in every division. There is a
! 20 branch of this LPCI function which is also heat removal, so 21 there are three complete divisions of RHR.
22 And one of the high pressure systems is the RCIC i
23 steam driven system which we've upgraded to safety grade to 24 make it more reliable and to count on it as an ECCS function
( 25 which we did not in our previous designs, i
Heritage Reporting Corporation (202) 678-4888 l
248 1 From the previous presentation today, you saw that we 2 basically got rid of the singla largest design basis accident 3 that we've analyzed in the past by eliminating the 1
4 recirculation piping. That has permitted us to rationalize 5 this system in having considerably lower capacity in both the 6 high pressure and in low pressure categories. No large pipes 7 located below the core.
8 A summary of what I'm going to tell you in a few 9 moments about peak clad temperature, basically we have no core 10 uncovery for any of the design basis accidents for ABWR. So 11 that in a very brief nutshell outlines the design.
12 (Slide) 13 This is another view of the ABWR emergency core
(])
14 cooling system high pressure. The RCIC is pretty much the same 15 as you are used to in previous BWRs. It is one of the three 16 divisions only. It is designed for 800 gallons per minute. It 17 operates as a normal makeup system in case of isolation or o 18 station blackout, for example, in addition to its ECCS duties.
19 It comes on at a higher water level than the other two high 20 pressure ECCS systems, so we have achieved separation of 21 initiation levels in this design.
22 (Slide) 23 The other two high pressure systems are called HPCF, 24 high pressure core flooder. There are the other two divisions 25 at rated conditions, that is to say at the safety valve set i
Heritage Reporting Corporation (202) 628-4888
() 1 point, they can produce 800 gpm, that is to match the RCIC flow 2 which is the short term need for the KE removal and act as a 3 backup for RCIC.
4 (Slide) 5 At low pressure, they provide, these are motor driven 6 pumps. They provide considerably more flow, over 3,000 gpm per 7 pump per division.
8 MR. EBERSOLE: How many diesel plants do you have 9 which are dedicated to this single load without cross 10 connections to other loads?
11 MR. SAWYER: Would you explain that again?
12 MR. EBERSOLE: Yes. You know on the intermediate
(} 13 designs, you had an independent diesel driven high pressure 14 core spray. That was the last design prior to this one. That 15 diesel was loaded only to that high pressure core egray pump.
16 MR. SAWYER: That's correct.
17 MR. EBERSOLE: In some designs, unfortunately, it had 18 a dependency in that the diesel cooling water was dependent on 19 the other diesels, which you fixed.
20 MR. SAWYER: Yes, sir.
21 MR. EBERSOLE: This new design, how many independent
- 22 AC power loops with only a singular load do you have?
23 MR. SAWYER: With only a single load?
24 MR. EBERSOLE: Yes.
25 MR. SAWYER: We don't have any.
l Heritage Reporting Corporation (202) 628-4888
250
~
1 MR. EBERSOLE: You don't have the high pressure core 2 spray anymore?
3 MR. SAWYER: That's correct. We have --
4 MR. EBERSOLE: Where does the electric power come i
5 from?
6 MR. SAWYER: The electric power for each of these 7 divisions comes from a diesel generator that supports these 8 loads as well as other emergency loads.
9 MR. EBERSOLE: So you've fallen back to the old 10 system of diesels which have a multiplicity of loads?
11 MR. SAWYER: They have more than one load on them.
12 What we have done is we have made a design change to increase
() 13 the amount of time required for diesels to start up.
words, we've relaxed the requirements on the diesels.
In other 14 ,
15 MR. EBERSOLE: But you haven't though parked any more 16 AC power as you did in the BWR\5s and \6s, you haven't parked 17 any AC power off by itself?
18 MR. SAWYER: We haven't parked any AC power off by 19 itself, that's correct.
20 MR. EBERSOLE: To that extent, you sre more dependent 21 now on ordinary diesel plant design than you were on the BWR\5 22 and \6?
23 MR. SAWYER: In some respects. In some respects, 24 what you're saf ing is right. We are going to address the o
k-) 25 potential core mold failure concern that comes from that by Heritage Reporting Corporation (202) 628-4888
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(')g 1 requiring that at least one of the diesels come from a 2 different procurement source.
3 MR. EBERSOLE: How many diesels are there?
4 MR. SAWYER: Three.
5 MR. EBERSOLE: Three. Okay.
6 MR. SAWYER: So we're aware of the concern that come-7 from having three identical diesels.
8- MR. EBERSOLE: Will you take other steps to seclude 9 it from the other set of two?
10 MR. SAWYER: There are no cross connections 11 whatcoever between the three divisions.
12 MR. EBERSOLE: The package and everything.
{} 13 14 MR. SAWYER:
MR. EBERSOLE:
The package is separate and everything.
Okay. So in a way, you do preserve 15 some aspects of the old \5 and \6.
16 MR. SAWYER: And the only remaining failure are what 17 you might consider to be design failures. And we're trying to 18 address that by requiring one of them to come from a different 19 source.
20 You will notice in those charts if you look carefully l 21 that some of the information was penned in. At our January 22 presentation and on all previous presentationr on the ABWR, 23- those two high pressure systems that were doing the injection 24 function into the vessel were called high pressure core spray.
() 25 I want the subcommittee to be aware of the fact that I
( Heritage Reporting Corporation I
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i 252 g3 V we have already notified the Staff that in our future i
2 submittals, as well as going back and correcting this 3 submittal, I'm not exactly sure what we've committed to for a 4 date, Joe, but it's mid-year or thereabouts, we are going to 5 delete the core spray sparger. That is to say, we are not 6 going to delete the function. The loop is there, the pump is 7 there, the diesel's there. Everything's there except the 8 sparger which is inside now, has a flooding function only 9 instead of a core spray function.
10 I want you to be aware also that in the Chapter 6 and 11 Chapter 15 analyses which we have already provided, we never 12 took credit for the spray function anyway as a cooling
(} 13 14 function. So the last bullet on here says that the design basis accident analyses will be essent ially unchanged from what i
15 have already been submitted.
16 The major reasons why we made this change are 17 twofold: First of all, EPRI ABRW requirements program, the 18 U.S. utility guidance to us, based on a lot of operating 19 experience was that they believed in tne interest of 20 simplification and ease of maintenance and surveillance, they l
21 wanted us to eliminate those functions unless we found them 22 necessary to meet our safety requirements. And we don't.
23 And because in our own evaluations we went back and 24 we agreed with them. We can simplify the maintenance and
( 25 surveillance of these spargers. The overhead spray sparger Heritage Reporting Corporation (202) 628-4888
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1 which we had in the design before would have had to have been 2 pulled out of the vessel every refueling outage. And was an 3 integral part of the shroud head and it was a bear to say the 4 least to conduct, would have been a bear to conduct in-service 5 inspection on it.
6 So I wanted the subcommittee to be aware that we've 7 made that change.
- 8 (Slide) 9 Low pressure core cooling systems. It is mostly 10 three divisional. Where it is not, I have parens with 2s for 11 the fuel pool cooling backup function which is rarely used 12 anyway. For the drywell spray function. And for the wetwell
(} 13 spray function, only two of the three divisii.na support those 14 functions.
15 The primary loops shown here are the ones that are 16 used that you would expect to be used most of the time. The 17 major one is suction from the pool, discharge to the vessel for 18 the LPCI function, suction from the vessel, back to the veesel 19 for shutdown cooling function.
20 One major improvement we have made over previous BWRs 21 is that there are separate shutdown cooling nozzles and 22 separate retu'rn lines for every division. So we now can 23 conduct the shutdown cooling function without worrying about 24 single failures losing that function.
() 25 HR. EBERSOLE: Well, do you partition those flows by Heritage Reportinn Corporation (202) 6 *i e 1888
}
254 O 1 flow meters to get concurrent sufficient flow cooling with just 2 a squeak of water to the vessel when you need it?
3 MR. SAWYER: Ah, let's talk about this. If you need 4 the system for suppression cooling, then it is dedicated to 5 suppression cool cooling. That is to say, this valve is closed 6 and that valve is open.
7 MR. EBERSOLE: You don't mix them?
8 MR. SAWYER: We do not mix them. There are three 9 divisions now.
10 Now, one other thing I forgot to mention and I almost 11 took this chart off, the heat exchanger shown here is always in 12 the loop. That is also a change from previous designs. That 13 is to say, whether you are conducting shutdown cooling, LPCI,
(")N w
14 or suppression cool cooling, that heat exchanger is always 15 present.
16 MR. EBERSOLE: One of three loops, it's got its own 17 AC power, its own diesel, its own DC and everything?
18 MR. SAWYER: Yes. Yes. Three loops.
19 MR. MICHELSON: You're not using the RHR in that so-20 called condensing mode any more?
21 MR. SAWYER: We do not. We eliminated the steam 22 condensing mode, and there were a couple of other add on modes 23 that the system we were required co do, one of which was the 24 vessel headspray which we've transferred to another system in Os_/ 25 order to simplify the performance of this system.
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255 0 1 MR. EBERSOLE: Do I see a valve to the containment 2 there that always kind of raises --
3 MR. SAWYER: That's the shutdown cooling valve.
4 -
MR. EBERSOLE: You have to open that although it's in 5 the containment?
6 MR. SAWYER: Yes. We have to open this valve. This 7 is the only part of the loop that is not counted on in 8 accidents. This is for normal shutdown cooling and you have an 9 inboard valve and an outboard valve. And if you have an 10 accident, this valve is not qualified for accident service. It 11 is normally closed. It wil' stay closed. We will assume that 12 we can't get it open.
13 MR. EBERSOLE: Where's the water going to go after it
(])
14 gets in the vessel?
15 MR. SAWYER: The water will go out the break.
16 MR. EBERSOLE: It's not much of a break.
17 MR. SAWYER: If it's not much of a break, then you 18 don't have much of a problem, right?
19 MR. EBERSOLE: I don't know it. You're not going to 20 cool it down a whole lot. You're going to keep on boiling.
21 You going to liquid cool it?
22 MR. SAWYER: No. We would use the relief valves to 23 blow down to the vessel.
24 MR. EBERSOLE: So you'll always be at caturation
( 25 boiling?
l Heritage Reporting Corporation (202) 628-4888
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1 MR. SAWYER: No. We'll bring the vessel all the way 2 down. Oh -- you're saying, what if I only have a small break?
3 MR. EBERSOLE: Yes.
4 MR. SAWYER: Then I will prevent boiling by using the 5 relief valve pathway.
6 MR. EBERSOLE: Okay. I thought you were going to say 7 that.
8 MR. SAWYER: Yes.
9 MR. EBERSOLE: You let the water go through them?
10 MR. SAWYER: Yes. Yes. For the small break case, 11 that's exactly how we solve that problem.
12 MR. EBERSOLE: There's enough of them and they're big
/~T 13 enough for that?
l kJ 14 MR. SAWYER: Yes. They have no problem passing this i 15 water flow at low pressure.
16 MR. EBERSOLE: So you're completely full of water.
17 MR. WADE: We'd go solid in that case.
18 MR. SAWYER: That's right.
19 MR. EBERSOLE: Eventually you could get at the core, i
i 20 you will depress it and reside on a low time factor to get it 21 open.
22 MR. SAWYER: Yes.
23 (Slide) l 24 I just have one summary chart. In the proprietary 25 section of 6.3, we provided you with accident performances for i
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r3 257 O 1 a wide spectrum of pipe breaks and for break locations and also 2 sizes of breaks for each of those locations. And a summary of 3 that is very simple.
4 First of all, no design basis accidents leads to core 5 uncovery. The limiting event is the break of one of the HPCF 6 lines itself. That leads to the lowest water level. 'That's 7 limiting in that sense.
8 I should point out to you that from a performance 9 standpoint, we in ord to avoid a bunch of complex arguments 10 and simplify both the d sign of the recirculation system as a 11 power supply system and not get into complicated licensing 12 arguments about whether we need to take credit for pump coast l
13 down, we conservatively analyzed our local performance by l t'])
14 tripping all the recirc pumps at T=0.
15 Even though when I stand here half an hour from now 16 and talk to you about all recirc pump trip, we consider that l
17 event to be an accident in its own right. That is why we L
18 calculated peak clad temperatures. It's not because of LOCA.
l 19 In other words, it's the fact that we tripped all the recirc 20 pumps at time zero is what leads to our calculating peak clad 21 temperatures at all. So I wanted you to be aware that this and 22 this are not coupled. They're uncoupled. It's because of that 23 assumption that we calculated peak clad temperature at all.
24 MR. EBERSOLE: Is this set of numbers up there 25 tantamount to saying, if I appropriately pressurize the i Heritage Reporting Corporation l (202) 628-4888 l
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() 1 reactor, but I maintain floor cover, and I don't have any 2 pumps, that's the temperature I'm going to get which is not 3 damaging?
4 MR. SAWYER: Not damaging. That's correct. We don't 5 expect to have any fuel damage from those.
6 MR. EBERSOLE: That's w'ithout pressure.
7 MR. SAWYER: This is without --
8 MR. EBERSOLE: Without positive pressure on the 9 water.
10 MR. SAWYER: You mean, in terms of net positive 11 suction heads for the water?
12 MR. EBERSOLE: Net positive pressure on top of the 13 water diminish the fraction? What pressure is that?
(])
14 MR. SAWYER: Some of these happened while, this 15 temperature happens within about 3 or 4 seconds of time zero, 16 mainly because it occurred from the momentary dryout th-17 occurs from the pump trip, ar.d you very quickly rewet.
18 MR. EBERSOLE: So that's just, well, that's at 19 containment pressure, isn't it? Or is it slightly more than 20 that?
21 MR. SAWYER: No. What I'm saying is that this 22 temperature occurs at the very beginning of the event while the 23 containment is still at low pressure, while the reactor is 24 still at high pressure. Later, very quickly, this goes back to 25 saturated conditions --
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1 MR. EBERSOLE: I'm trying to get a fix on temperature 2 at loss of pressure in the coolant?
3 MR. SAWYER: Very quickly, very quickly after this, 4 the clad rewets and from then on it tracks saturation 5 temperature all the way down.
6 MR. EBERSOLE: Not bothered by voids?
7 MR. SAWYER: Not bothered by voids.
8 MR. EBERSOLE: At low pressure.
9 MR. SAWYER: Not bothered by voids at low pressure.
10 MR. EBERSOLE: Okay.
11 MR. SAWYER: That's correct. Whether it's boiling or 12 it's not boiling at lower pressure will have no affect on it.
13 MR. EBERSOLE: See, we've long wanted to be able to
(]')
14 make the general statement, as long as I have water cover, 15 irrespective of content, I'm not going to have damaging 16 temperatures.
17 MR. SAWYER: That's correct.
18 MR. EBERSOLE: That's a very broad nice thing to say.
19 All I need is water level.
20 MR. SAWYER: All you need is water level. That is 21 absolutely right.
l 22 MR. EBERSOLE: A water level or the knowledge of 23 having it or where it is is critical to safety. Where is it?
l 24 MR. SAWYER: That's correct.
n
- 25 MR. EBERSOLE: Nothing else, just water level.
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, 260 1 MR. SAWYER: Just water level. In,this case, I need 2 some front end assist to replace the water I lost from the 3 break, but after that, it's all over.
4 MR. EBERSOLE: See, it's nice to be able to say, I 5 don't need any parameter but water level to tell me where I am 6 if I tip the reactor.
7 MR. SAWYER: Yes. In the short term, you're 8 absolutely right, Jesse. In the long term, I would add to that 9 that we need to know something about the pool as a measure of 10 the containment performance.
11 MR. EBERSOLE: Oh, sure.
12 MR. SAWYER: Okay. That's Section 6.3. If there
(} 13 aren't any questions in this area, I'll proceed to Chapter 15.
14 What you saw was sort of as a preview of upcoming 15 attractions. As you know, we are in the process of doing our 16 Chapter 15 PRA. I guess it's going to be called Chapter 19 17 this time around.
18 And I think Dr. Kerr wanted to know had we used the 19 PRA process to guide the design and how sure were we that we 20 had at this point that we had the right features in there.
21 And I want you to be aware -- let me say some caveats 22 first, and then draw some conclusions from this chart.
23 (Slide) 24 The caveats are that in our submittal in December, 25 the ABWR numbers are almost certain to be different than what's Heritage Reporting Corporation (202) 628-4888
261 7s .
'^' )'
1 here just because the ground rules are different.
2 This was an analysis that I did myself with hand 3 calculations without sophisticated PRA techniques, but using 4 the kind of fault trees that we're familiar with from previous 5 PRAs and it's basis, the main reason for doing this was to 6 compare current BWRs and the ABWR on an apples and apples basis 7 as much as possible.
8 That is to say, I used the same frequencies of 9 occurrence for transients, for LOCAs, for loss of offsite 10 power. I used the same equipment reliability if the various 11 product line had the same equipment. And the purposo of 12 generating this chart was to say, have I improved things with my ABWR safety.
{) 13 14 And I categorized them in many categories: loss of 15 offsite power, transients, LOCAS, loss of decay heat removal, 16 ATWS and total.
17 MR. EBERSOLE: This is without UPBS?
18 MR. SAWYER: This is without UPBS. That's right.
19 The BWR\6 does not have UPBS.
20 (Slide) 21 MR. EBERSOLE: With UPBS, where would we be?
22 MR. SAWYER: With UPBS, based on our BWR\6 studies, 23 it will reduce the core damage frequency by about a factor of 24 ten. And it will reduce several categories. It will reduce l
( 25 this one. It'll also reduce that one. And that one, too.
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262 F) 1 MR. EBERSOLE: Well, if it's so cheap, Craig, why 2 don't we have it?
3 You don't think that factor of ten is worth it?
4 MR. SAWYER: Well, that's the point. It gets to be 5 -- there is a point where you have to say, when have I got' 6 enough equipment in this place, okay.
7 MR. EBERSOLE: I know. F:ut you know, that has such 8 an appeal because the common man can understand it.
9 MR. SAWYER: Well, I'can tell you this we will be 10 studying that option as part of our PRA to come to some 11 determination as to whether the cost benefit makes sense.
12 MR. EBERSOLE: Look at it in the context of e
13 salability to the party who doesn't know a hell of a lot about
(]}
14 these complex systems. You know, public acceptance. It may be 15 worth a lot there, whether it's worth technically a lot.
~ 16 Okay?
17 MR. SAWYER: Yes, I understand, yes.
l 18 MR. EBERSOLE: You know, like.in Shoreham.
19 MR. SAWYER: Right. I'm not sure that would have 20 saved Shoreham, however.
21 MR. EBERSOLE: I don't think so either.
22 MR. SAWYER: Okay, let me give you a quick overview 23 of Chapter 15. And what we've done is prepare the chapter per 24 the Reg I.170 format. And it covers transients, even though it N/ 25 says accidents, basically it covers what we call, transients, Heritage Reporting Corporation (202) 628-4888 t
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(> 1 as well as accidents.
2 The presentation is packaged into several packages:
3 decrease in coolant temperature; increase in pressure; decrease 4 in flow' rate; reactivity' anomalies; increase in inventory; 5 decrease in inventory; and ATWS. We did this to meet the 6 format guidelines.
7 However, in the rest of my presentation as I 8 summarize here, I will combine things where it makes sense for 9 your ease of understanding.
10 The other thing that we have provided in the Chapter 11 15 submittal today is a nuclear safety operational analysis, 12 and I'll have some words to say about that as I get there.
(]} 13 (Slide) 14 Before I talk about transients, let me go back for a 15 few minutes and talk about fuel design as presented in Chapter 16 4 and in the G Star submittal. As Irv Kobsa mentioned, we're 17 going to be using standard fuel designs. Our SAR analysis that 18 we provide in Chapter 15 is based basically on 8x8 barrier 19 fuel. That's the basis.
20 The steady state core configuration which was the 21 basis of the starting point of our analysis was based on a 22 control self-core fuel management scheme which is pretty much 23 the standard that GE is offering today to our utilities. One 24 design feature which has already been covered, and that is that 25 we did increase the control rod pitch slightly from our Heritage Reporting Corporation (202) 628-4888
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',b i 1 previous product and lines primarily to reduce the void l 2 coefficient and get, better transient performance.
3 In terms of steady state performance, there are 4 several ways to look at it. Minimum critical power ratio is 5 the measure we use in BWR terminology to measure how close we !
1 6 we are to departure from a nuclear boiling.
7 Maximum linear heat rate which we quote in kilowatts 8 per Ecot is a way of characterizing our fuel mechanical 9 performance and what we're willing to warrant in the way of 10 good fuel performance.
11 And the other thing that you worry about primarily in 12 fuel design is how am I doing, am I hot to cold. That is to 13 say, if we're giving today, how am I doing cold.
(]}
14 I'm going to be showing you a lot about how the 15 transients that we've analyzed support that the operating limit 16 CPR is about 116 which compares to about 119 or 120 or 17 thereabouts on current product lines. So the message we have 18 to say basically is that we think that this product has quite a 19 bit of margin in it between steady state operation and the 20 limiting additions for operation.
21 The same thing's true in the linear heat generation 22 category. The limit's 14.4 and we don't see from our steady 23 state analyses any time where the reactor will go over 11. And 24 similarly, a considerable amount of margin for cold shutdown.
A V 25 (Slide)
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265 1 For those of you who don't know what control' cell 2- core means, I took the Chapter 4.3 figure and read it in the 3 low enrichment initial core, which will be replaced by highest 4 exposure fuel in reload cores of the 13 control cell core 5 locations. Basically, we designed this reactor for a minimum 6 hot access and nice flat operating hot access as a function of 7 exposure so that only the maximum number of control rods needed 8 at power is 13. No rod swaps.
9 So we've designed this thing for a much simplified 10 fuel management relative to what's been in the past. I'll give 11 you a minute to shade it in.
12 Transient performance. There are some things that 13 you don't know about yet because we haven't supplied Chapter 7
(])
14 and Chapter 8, so I'm going to have to tell you about them 15 today. And then you get to review them when our Chapters 7 and 16 8 submittals really come in.
17 In the approach to startup and power range operation, 18 we have incorporated some new design features. In the startup 19 range, we've gone with wide range neutron monitor systems which 20 replace the combined source range monitor and intermediate 21 range monitor and have a period based protection system. It 22 isn't based on period done as a derivative; it's based on 23 period derived as an integral function, and it amounts to the I
24 same thing except it gets rid of the noise.
I 25 And that feature is what replaced the upscale range Heritage Reporting Corporation (202) 628-4888
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1 switches.
2 MR. EBERSOLE: You mean this is an integral which is 3 integrated on a period basis?
4 MR. SAWYER: Jes, sir. The flux is basically time 5 integrated and that --
6 MR. EBERSOLE: What's the panel span, do you know?
7 MR. SAWYER: I think the sampling period for this is 8 the order of a second or half second, something like that.
9 MR. EBERSOLE: That gets rid of the noise problem?
10 MR. SAWYEP: Gets rid of the noise problem, that's 11 right.
I 12 In the power range, the existing plants have rod
() 13 block monitors. We have a rod block monitor system too. We 14 call it an advanced rod block monitor system but one of the 15 features it does is it incorporates a rod block before you even 16 go below the operating limit. And in so doing, we are 17 justifying eliminate the rod withdrawal error as a transient.
18 IN other words, it's going to take multiple failures'of these 19 redundant systems in order for us to have a rod withdrawal 20 error at least analyzed as a transient.
21 Analyzed as an accident, okay, but no longer as a 22 transient.
23 Similarly in the control system area, we've gone from j 24 our traditional one-channel analog systems to three channel l
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1 the control systems signal failure proof. So rather than the 2 analyses which we provided in Chapter 15 in the control system 3 failure area, concentrate on failures of single pieces of 4 equipment as opposed to failing the entire system which we've 5 done in the past.
6 This just gives you a preview of some expectation of 7 gentler transients. I mentioned that some of the design 8 changes we've done have gone in the direction of reducing the 9 void coefficient, we have more water in the vessel so that if 10 one measures this thing we call level rate which is the rate at 11 which you would lose level if you instantaneously turned off 12 all incoming water, and the level rate is gentler. And the l
13 pressure rate has been reduced primarily because of the void
({)
14 coefficient.
15 (Slide) 16 What those things lead to is critical power 17 performance in the various categories. Decrease coolant 18 temperature, increase reactor pressure, decrease coolant flow 19 rate, etcetera, etcetera. If you go down this list, there's a 20 summary table in Section 15.0 that summarizes on one great big l
21 matrix the results of all these transients.
22 And if you go stand those tables, you'll find in each 22 of these categories, there's one event that limits that l 24 category, and then if you look at the worst of all of those, O 25 you'd find that .09 is the largest delta CPR event. So if you Heritage Reporting Corporation (202) 628-4888
() 1 put that together with the statistical evaluation which we do 2 which sets what we call the safoty limit, which means that in 3 order to assure that less than a tenth of a percent of rods are 4 in transition boiling, I have to assure that my minimum CPR 5 statistically must be above 107. And on top of that, I've got 6 to add another 09 to accommodate the~ worst transient, so that 7 sets my operating limit at 116.
8 So that's how the 116 was derived that showed up a 9 couple pages ago on .the fuel chart.
10 (Slide) 11 In terms of over pressure perfo'rmance, we went into 12 this a little bit, earlier. The requirement for overpressure
() 13 is to be less than ten percent above the design pressure for 14 overpressure events. That's where the 1375 psig comes from.
15 If you look at the various pressurization transients, not from 16 a CPR point of view, but from a pressurization point of view, 17 you get the following kind of list.
18 And the limiting event on this list is the MSIV 19 closure back-up scram which is what we call the ASME event.
20 And that calculatas 1274 with 1375 being the peak primary 21 pressure.
22 (Slide) 23 LOCAs we covered in Chapter 6. Let's talk about 1
! 24 other events.
() 25 MR. EBERSOLE: Well now wait a minute. MSIV closure Heritage Reporting Corporation (202) 628-4888
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C:) was normal scram, wasn't it?
I 2 MR. SAWYER: This one here is not the event we 3 analyzed for transient MCPR compliance. This is the event we 4 analyzed for compliance to over pressure criteria. It's MSIV 5 closure, position scram failed, flux scram.
6 MR. EBERSOLE: But it is with scram.
8 MR. EBERSOLE: Now, the thing we were asking about, 9 if you recall, was it without the scram. Can you --
10 MR. SAWYER: Yes, we'll get to that. I've got an 11 ATWS presentation to give you in a little bit.
12 This right now is just everything but ATWS.
({} 13 (Slide) j 14 In accident performance, the LOCAs were covered in 1
15 Chapter 6, which we talked about previously. We will be 16 providing you with a detailed electrical network for the 17 recirculation pump power supply design. But basically, we have
~
18 provided it, we've taken our ten p':mps and divided them 2,?;
19 2,3, with different power supply sources for them. So that it 20 takes multiple failures.for you to postulate simultaneous loss 21 of power to all the pumps.
22 As a result, we are categorizing an all recirculation 23 pump trip as an accident event and it quite naturally creates 24 peak load temperatures similar to the ones you saw for LOCAs I
("N l k/ 25 because those were front end driven by this assumption anyway.
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1 In terms of reactivity accidents, the two that 2 people most worry about are the rod drop and rod ejection.
3 Now, you notice from today's presentation, we have 4- justification in the hardware design to not need to analyze 5 these events at all because it takes many failures for a rod 6 drop accident to occur, and many failures for a rod ejection 7 accident to occur.
8 Nonetheless, in a submittal which we provided the 9 Staff earlier last year, we did do an analysis anyway, and it 10 basically gave very minor results. The criteria for success on 11 these is 280 calories per gram. So these are really not, these 12 events, these kind of events have become, no, never minds for 13 BWRs.
(])
14 Similarly, as you might imagine since we've gone from 15 two pumps to ten pumps, pump seizure in any individual pumps is 16 really not going to cause much of a problem from a transient 17 point of view any more.
18 The radiological evaluation is also provided for 19 several of these events in Chapter 15. For the reactivity 20 accidents, we provided a very bounding analysis assuming 280 21 calories per gram was given to the peak pellet even though our 22 analysis showed less than 100 calories per gram. And even if 23 we made that assumption, we calculated a very small fraction of 24 10 CFR 100 at the site dump.
25 MR. EBERSOLE: How did that get out?
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1 MF 3WYER: What?
2 MR. EBERSOLE: How did'that stuff get out? To 3 present itself to the public?
4 MR. SAWYER: It gets out through leakage through the 5 NSIVs.
6 MR. EBERSOLE: Okay.
7 MR. SAWYER: In the case of LOCAs, the thing on this 8 whole chart, the only thing of importance for you to note is 9 this double star footnote here where I say we are taking 10 significant deviations from Reg Guide 1.3, and are basically 11 following the guidelines of NUREG 1169. I'm sure this is going 12 to be a matter of continued discussion between us and the
[} 13 Staff. But that's the process that we're using for doing our 14 DBA evaluations.
15 As a result of that, we've been able to relax our 16 leakage criteria on the MSIVs and still get a quite acceptably 17 low doses both at the low population zone and at the site 18 boundary compared with 10 CFR 100 limits.
19 Okay. Before going on to NSOA, let's talk about ATWS 20 for a few minutes.
21 Our FMCRDs as we've described to you have to diverse 22 insertion loads: the hydraulic scram which is very rapid, about 23 a second and a half or so; and electric run in which takes 24 approximately three minutes from full out to full in. The l
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-~ 272 1 signals. That's to say, they come from different sources.
2 The RPS system supplies the signals for the hydraulic 3 scram, whereas the rod control information system supplies the 4 electric run in signals so we believe with this drive we have 5 addressed the previous common mode failure concerns that have 6 been identified by the Staff in coming to grips with how and 7 how much you need to have and wha +. you need to have to address 8 ATWS.
9 We do have recirculation pump trip based on high 10 pressure and or low water level to limit the peak pressure.
11 That is consistent with saying, what if the hydraulic scram l
l 12 failed and I need my electric run in backup. How do I handle l
13 over pressure in the meantime, and the answer is, my
({}
14 recirculation pump trip limits my overpressure.
15 Our positi-on is that we don't need the SLCS for ATWS 16 due to the other features that we are supplying. And we're 17 going to have on-going discussion with the Staff about our 18 compliance with 5062. Unfortunately, 5062 was very 19 prescriptive and written around existing drive designs. It 20 doesn't give us much room to maneuver.
21 MR. EBERSOLE: Does the manual kick filter require 1
22 coincident actions?
23 MR. SAWYER: Manual capability exists and you can --
24 MR. EBERSOLE: Yes, but does it preclude incidental
( 25 operation by requiring coincident, you know, two valves or I
l l
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1 whatever it takes to really do it.
2 MR. SAWYER: It takes one valve to open.
3 MR. EBERSOLE: Would you want to put two?
4 MR. SAWYER: No. It only takes one out of two.
5 There are two. In other words, the SLCS system itself is two 6 divisional --
7 MR. EBERSOLE: But one mistake though will inject it, 8 won't it? Do you think, one mistake will inject it?
9 MR. SAWYER: One mistake could inject SLC.
10 MR. EBERSOLE: I would argue you need two valves to 11 prevent one mistake injection.
12 MR. SAWYER: We have addressed 'onat to some extent,
() 13 Jesse, in the test design of a system. We designed that thing 14 to be tested with clean water rather than boiling water to 15 address that kind of issue.
16 MR. EBERSOLE: Well, it wouldn't be a hazard to have 17 coincident valve requirements.
18 MR. SAWYER: We have to look at that from a 19 probabilistic standpoint because you want to have that function 20 available on demand, too.
21 MR. EBERSOLE: Sure. Yes, but it'd cost you 22 something to accidentally do it.
23 MR. SAWYER: Okay. There's some conflict between 24 5062 and Reg Guide 170 and as a result, we didn't submit a 25 formal analysis showing our ATWS performance. But since you Heritage Reporting Corporation (202) 628-4888
V g 274
(_/
1 asked the question, I can go b.ack to some analyses that we have 2 done, although we haven't documented them to the Staff, for the 3 worst ATWS which is MSIV closure. That's the one that gives 4 you the highest peak pressure. The peak pressure we calculate 5 is 1340 psig, which is less than the Service Level B criteria.
6 We've done that on purpose, even though we think 7 we're entitled to Service Level C for ATWS, mainly because we 8 want to simplify the design of the plant from a qualifications 9 standpoint.
10 This should have had a less than, I see I forgot to 11 put that in there, less than 8 percent rods in transition 12 boiling. A peak pool temperature of about less than 160 13 degrees F. and these kinds of numbers are well within the
(])
14 previous studies that we've done, and the Staff has done and 15 other people have done which led to the formulation of 10 CFR 16 50.62 in the first place.
17 (Slide) 18 Finally, I'm not exactly sure how to present this 19 subject because there are many many many flow charts in this 20 section of Chapter 15. The reason why we at GE started doing 21 these NSOAs, we started these years ago before there were any 22 formal FMEAs being done.
23 And the reason we started doing them was basically to 24 identify what were the safety functions required and to 25 demonstrate compliance of the single failure criteria. And Heritage Reporting Corporation (202) 628-4888
275 7-)
V 1 that is, we would make a list of how to, you know, reactivity 2 control, decay heat removal, core cooling, containment 3 performance. And then we'd go down that list for each one of 4 the events and figure out what equipment was required to 5 achieve success including a single failure.
6 So it's sort of like an FMEA but not exactly. And so 7 the process that we go through is we define the events,'and as 8 I said here, we went through 56 individual kinds of events. We 9 went through all of the kinds of transients and accidents we've 10 ever considered at the BWR. We defined what constitutes 11 acceptance criteria for each event.
! 12 We then go ahead and define, that is to say, this is
(')
ss 13 like 2200 degrees of Peak clad temperature for 280 calories per 14 gram, or MCPR greater than safety limit, that's acceptance 15 criteria.
16 Then we go ahead and define success criteria. How 17 many systems does it take in order for this function to be 18 successful. We included all the plant states. Not just full 19 power operations. We included start-up, refueling, and all 20 areas of plant operation.
21 MR. EBERSOLE: Did that include the hypothetical 22 failure of some service system?
23 MR. SAWYER: Service systems are treated in there as 24 a column mode support state and if you look in the back of i
(j 25 Section 15.b, there's about half a dozen charts. Events i
Heritage Reporting Corporation (202) 628-4888 l
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- 276 1 through the special analysis of the capability or the 2 vulnerability if you will of the support systems.
3 MR. EBERSOLE: Well, if you fail a needed system, and 4 you only have one system to back it up, and it suffers a 5 somewhat significant increase in demand, do you then invoke 6 thesis it may fail too, and therefore you already have lost 7 your single failure criteria?
8 MR. SAWYER: We didn't analyze it that way. In other 9 words, when developing our success criteria here, as opposed to 10 the PRA, success criteria here were defined'in the licensing 11 context. In other words, let me give you an example. In the 12 PRA, we would say that any one pump constitutes success for 13 core pooling. In the design basis world, we'd eay, it turns 14 out we need two pumps with the worst single act of failure.
15 In the PRA, we'd say we only need one heat removal 16 loop. In the licensing world, we'll say two. Here we used 17 licensing success criteria.
18 MR. EBERSOLE: What I'm really getting at is if you 19 stick to the thesis that given an accident which might include 20 the failure of a service system, do you require then that you 21 can suffer a single random failure and still survive?
22 MR. SAWYER: Of course. That's part of this 23 analysis. We do take inte account the single act of failure 24 for the events that require that kind of thinking.
Okay.
25 MR. EBERSOLE: Great.
Heritage Reporting Corporation (202) 628-4888
277 (0S '
1 MR. SAWYER: Yes. That's in those charts, and 2 .there's probably 70 pages of such charts and there's.a lot of 3 detail to go through.
4 MR. EBERSOLE: That ought to make some interesting 5 reading.
6 MR. SAWYER: Yes. So that in a nutshell is what's in 7 Chapter 15. I think we had some discussion this morning that 8 you will be receiving a Section 15.B later this year that will 9 go into some more things you're interested in, specifically for 10 the FMCRD and the recirculation pumps on failure modes and 11 effects for them.
12 MR. MICHELSON: When will the severe accident and
() 13 other types of considerations appear?
14 MR. SAWYER: Chaptor 19 as part of the PRA.
15 MR. MICHELSON: You'll consider that all a part of a 16 PRA?
17 MR. SAWYER: That's right.
18 MR. MICHELSON: Hcw about the section, where will we 19 discuss the compliance with USIs and GIs and so forth?
20 MR. SAWYER: Is that in Chapter 3, compliance with 21 USIs and GIs?
22 MR. CORK: We are dealing with that right now, I 23 know. I'm trying to recall the date of our submittal. I'm not 24 sure. Dino, do you know?
O 25 (Discussion trying to locate information off the Heritage Reporting Corporation (202) 628-4888
278 3
1 record.)
2 MR. MICHELSON: Some of these GIs and USIs deal with 3 problems that perhaps need to be taken into consideration when 4 doing a chapter 15 kind of analysis. And I just wondered how 5 you splayed the two together since you don't put it in Chapter 6 15 and you put it in Chapter 3, then somehow you have to lead 7 the reader to believe what affect this has on the Chapter 15 8 analysis. Or how you have to change it or whatever.
9 Systems interaction under 817 and 847 clearly can 10 affect the answers you arrived at in Chapter 15 or whether or 11 not you've included system interaction af fects in doir.g the 12 analysis. And maybe you put this in a severe accident chapter
() 13 which could also be done, or you can consider these USIs and GI 14 resolutions of a severe accident --
15 MR. SAWYER: Well, it's required by the severe 16 accident policy statement to be addressed.
17 MR. MICHELSON: Yes, I know.
18 MR. SAWYER: And that's why I surmised that we'd 19 probably be submitting it as part of that PRA.
20 MR. MICHELSON: So Chapter 15 appeared to be what I'd 21 call the classical chapter. That's the stuff you've seen 22 before the standard accident you rolled through, and so forth.
23 but it doesn't include such things as the effects of pipe 24 breaks outside a containment, which are accidents, and they're O 25 not necessarily severe although they might get into that Heritage Reporting Corporation (202) 628-4888
l i
()
1 category.
279 i
2 MR. SAWYER: Especially from a radio. logical 3 standpoint.
4 MR. MICHELSON: Beg your pardon?
5 MR. SAWYER: We've-analyzed that from a radiological 6 standpoint.
7 MR. MICHELSON: I don't have any problem with the 8 radiological standpoint. I have a problem with the 9 environmental effects of such experiences, the flooding 10 effects, the straining effects, the steaming of compartments 11 pressurization of areas, things of this sort.
12 And normally, we haven't thoughc of those as severe
() 13 accidents because they were after all required to be addressed 14 way back in 1972. And we didn't call them severe accidents 15 then. We just wrote a letter, the NRC wrote a letter at that 16 time and said you '.all go through an analysis and let them 17 know about these pipe breaks.
18 But they have not traditionally appeared in Chapter 19 15. And I would expect them to appear somewhere but I'm not 20 quite sure where.
21 MR. SAWYER: Gentry, don't you have a department 22 analysis for high energy line breaks in Chapter 6.2?
23 MR. WADE: We don't have any results in yet.
24 MR. SAWYER: Okay.
'- 25 MR. MICHELSON: That would be compartment Heritage Reporting Corporation (202) 628-4888
280 0 1 pressurization is one of the phenomenon, but another one is 2 what happens when you expose the equipment to severe 3 environments for which they may or may not be qualified.
4 Flooding effects, submergence, things of this sort. In other 5 words, where do your pipe break analyses outside of containment 6 appear?
7 MR. SAWYER: That's a good question. I don't know 8 the answer. I know that we owe it to you because it was 9 something we owa you on the environmental stuff, you know.
10 What are the temperatures and pressures for normal operation 11 and abnormal operation parts of the reactor building and i
12 containment. And I know that we're on a schedule to supply 13 that but I don't recall what chapter that's in.
(])
14 MR. MICHELSON: We'll see it eventually. I would 15 expect either to see it or see an explanation of why it isn't 16 needed, either way.
17 I believe that that completes the GE commitments to 18 the meeting. Is that correct? Do you have any other 19 statements that you wish to make at this time?
20 MR. CORK: No.
I 21 MR. MICHELSON: The staff give us a few minutes for l
l 22 their views?
23 MR. SCALETTI: We are very early in the review 24 process for the ABWR. Unfortunately many of the areas of l
l 0
%- 25 concern that are topics of discussion today were centered i
l Heritage Reporting Corporation l (202) 628-4888 i
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281 1 around Chapters 4 and 5, and those questions have yet to be 2 generated by the Staff, and are.in the process of being 3 generated, and it is due to be given to G.E. by the end of this 4 month.
5 We did have some internal adjustments to the schedule 6 to take care of certain resource constraints that we had, will 7 not impact the overall schedule, so we'll be back on schedule 8 internally by December of this year. And the SCR should be 9 published on schedule.
10 That's all I have to say.
11 MR. MICHELSON: If you can wait just a minute then, 12 I'll address the subcommittee. I will make a presentation to
(} 13 the full committee, just a subcommittee report I believe on 14 Friday, it was, Friday at 2:30. The staff may certainly be 15 here if they wish. I don't know that it's necessary but it's 16 certainly not prohibited.
17 GE could be here if they wish although we don't 18 expect that you'll have to make any support.
19 The next subcommittee meeting is tentatively planned 20 for November of this year, before which time I will be sending 21 you an agenda for the next meeting. And that agenda now I 22 would anticipate it would be built upon two things.
23 First of all, we'll do the overview of the second 24 module, which are Chapters 1, 2, and 3. And then there will be
! f- 25 another part of the meeting devoted to follow up clean up l
Heritage Reporting Corporation (202) 628-4888 l
C
282 V,s 1 questions, areas that we want to explore further.
2 Those we'll try insofar as possible to identify to 3 you well ahead of time. They will be more pointed types of 4 questions but not any more broad gauge presentations. We've 5 received all the overview I think we need on these sections, 6 but members will have particular areas they'd like to hear more 7 about or particular quer .ons, and I'll put those together 8 with Richard as an agenda which I would hope we'd get out to 9 you well before the meeting. Like I'll try to get it to you 10 six to eight weeks before the meeting, if we can. So you'll 11 know what things we'd like you to concentrate on.
12 The overview though for Module 2 will be just like 13 the kind we had today. Again, I will try to tell you ahead of
({}
14 time which areas I think you want to focus mora on because 15 there's too much material to cover.
16 Now, the meeting we're hoping will still be a one-day 17 meeting but that depends upon how many follow-up things we 18 still have from today. If it's too many follow-up things, then 19 we'd like to work out a schedula for a part of one day and plus 20 a full day.
i 21 MR. CORK: Mr. Michelson, I'd like to second and 22 encourage your comment exchanging agendas at an early time. As 23 you can appreciate, a lot of our staff is tied up with 24 commitments in Japan, and so that if we could work out the fi 25 areas of interest to you, they could adjust their schedules so Heritage Reporting Corporation (202) 628-4888 l
i
- 283 1 that they could be here.
2~ MR. MICHELSON: Well, I appreciate that, and I really 3 honestly do not expect you to bring the people with all the 4 answers. I don't mind carrying over those items that we can't 5 get answers for at a particular meeting because from now on, 6 every module will get two questions.
7 You've got a copy I think of the schedule I handed 8 out?
9 MR. CORK: Yes. I do.
10 MR. MICHELSON: And you'll notice that except for 11 this first module, all the other modules will have a two 12 setsions in which we can ask follow up questions yet. And so 13 we don't expem' every answer on the first go around. We expect
(])
14 to follow up for clarification, and that's particularly 15 intended so ycu don't have to bring too many people with you.
16 We think this will work out fine, but we will make a very 17 honest effort to get the questions to you well ahead of time.
18 MR. CORK: Thank you.
19 MR. MICHELSON: Are there any subcommittee member 20 comments?
21 We do appreciate very much the fine support that 22 General Electric offered and the support of the Staff, and I 23 think this was a very profitable meeting, from our viewpeint at 24 least. And we will certainly be looking forward to following
' 25 up further in November of this year.
Heritage Reporting Corporation (202) 628-4888
. . ~ . ~ . . . , . ._. . . ..- ._
284 h
1 So thank you very much.
2 And I believe that finishes the meeting.
3 (Whereupon, at'5:00 p.m., the meeti:ig was concluded.)
-4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 l
20 i
21 j 22 l 23 1
24 25 i
Heritage Reporting Corporatiori (202) 628-4888
1 CERTIFICATE
/~T .
L~) 2 3 This is to certify that the attached proceedings before the 4 United States Nuclear Regulatory Commission in the matter of:
5 Name: SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR 6
7 Docket Number:
8 Place: Washington, D.C.
. 9 Date: June 1, 1988 10 were held as herein appears, and that this is the original 11 transcript thereof for the file of the United States Nuclear 12 Regulatory Commission taken stenographically by me and, 13 thereafter reduced to typewriting by me or under the direction 14 of the ccurt reporting company, and that the transcript is a D)
(_ 15 true and accurate record of the foregoing proceedings.
16 /S/ com N 17 (Signature typed): Joan Rose 18 Official Reporter 19 Heritage Reporting Corporation 20 21 22 23 24 25
%)
{)
s Heritage Reporting Corporation (202) 628-4888
INTRODUCTORY STATEMENT BY THE CHAIRMAN OF THE ACRS SUBCOMMITTEE ON ADVANCED BOILING WATER REACTOR, O ause 1. 1988. w^s"tacto". o c-THE MEETING WILL NOW COME TO ORDER. THIS IS A MEETING 0F THE ACRS SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR.
1 AM C. MICHELSON, CHAIRMAN OF THE ACRS SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR. THE OTHER ACRS MEMBERS IN ATTENDANCE ARE W.
KERR, F. REMICK, P. SHEWMON, D. WARD AND C. WYLIE. THE CONSULTANT PRESENT J. C. EBERS0LE.
THE SUBCOMMITTEE WILL BEING THE FINAL DESIGN APPROVAL (FDA) REVIEW 0F THIS STANDARD PLANT CONCEPT. THIS MEETING WILL CONCENTRATE ON THE FIRST REVIEW MODULE CONSISTING 0F SAFETY ANALYSIS REPORT, CHAPTERS 4, 5, 6, and 15 - 1.
RICHARD MAJOR IS THE COGNIZANT ACRS STAFF MEMBER FOR TODAY'S MEETING.
THE RULES FOR PARTICIPATION IN TODAY'S MEETING HAVE BEEN ANN 0UNCED AS PART OF THE NOTICE OF THIS MEETING THAT WAS PUBLISHED IN THE FEDERAL REGISTER ON MAY 18, 1988.
THIS MEETING IS BEING CONDUCTED IN ACCORDANCE WITH THE PROVISIONS OF THE FEDERAL ADVISORY COMMITTEE ACT AND THE GOVERNMENT IN THE SUNSHINE ACT.
WE HAVE RECEIVED NO WRITTEN OR ORAL STATEMENTS FROM MEMBERS OF THE PUBLIC.
IT IS REQUESTED THAT EACH SPEAKER FIRST IDENTIFY HIMSELF OR HERSELF AND SPEAK WITH SUFFICIENT CLARITY AND VCG ME S0 THAT HE OR SHE CAN BE READILY HEARD.
DO ANY SUBCOMMITTEE MEMBERS HAVE INITIAL COMMENTS AT THIS TIME?
(CHAIRMAN'S COMMENTS, IF ANY)
WE WILL N0W PROCEED WITH THE MEETING.
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PRESENTED TO ACRS SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR O
JUNE 1, 1988 WASHINGTON, D.C.
GE NUCLEAR ENERGY O
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O AGENDA i l
0 ABWR BACKGROUND O ABWR FEATURES l
l 0 RELATIONSHIP TO PROJECT IN JAPAN l
O O COMMISSION BRIEFINGS AND ACRS MEETINGS ,
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!' ABWR BACKGROUND O
! O ABWR DEVELOPMENT l
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0 PROGRAM OBJECTIVES MEET EPRI REQUIREMENTS O -
CERTIFY INTERNATIONAL ABWR DESIGN
~ "
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RELATIONSHIP TO PROJECT IN JAPAN 0 KASHIWAZAKI 6 & 7 SCHEDULE LICENSING APPLICATION 1988 K-6 COMMERCIAL OPERATION 1996 K-7 COMMERCIAL OPERATION 1998 0 GE/HITACHI/TOSHIBA JOINT VENTURE
() .
GE TO SUPPLY NUCLEAR STEAN SUPPLY, FUEL, TURBINE GENERATORS 9
0 U.S./ JAPANESE REGULATORY INTERACTION C)
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O O O ABWR Licensing Scheduies 1987 1988 1989 1990 1991 1992 I I I I I I I Japan Appiic ti nrr Establishment K-6 Construction Permit (EP) Rece.ipt of EP g
A A AA MITI Construction Permit issued U.S.
Licensing Review Basis issued ACRS Letter NRC Certification NRC ^^ ^ O ^
_ Final Design Approval
_ Application issued for Coftification i I I I I I i 1087 1988 1989 1990 1991 1992
O COMISSION BRIEFINGS AND ACRS MEETINGS 0 COMMISSION BRIEFINGS SEPTEMBER 1986 APRIL 1987 JANUARY 1988 O ACRS FULL COMITTEE MEETINGS JANUARY 1987 MARCH 1987 JANUARY 1988 0 ACRS SUBCOMITTEE MEETINGS JANUARY 1987 JUNE 1988 O ,
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O STATUS OF CERTIFICATION PROGRAM i
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SCHEDULE O _
REVIEW ALLOWS FOR MODULAR SUBMITTALS LEVEL OF DESIGN DETAIL ACRS REVIEWS .,
DESIGN CERTIFICATION PROCESS 0 KEY ACCEPTANCE CRITERIA PRA METHODOLOGY CORE DAMAGE <10-5/ YEAR FREQUENCY OF EXCEEDING 25 REM <10-6/ YEAR ADVANCED ELECTRONICS DESIGN GUIDELINES O
/88 l
STATUS OF CERTIFICATION PRGGRAM (CONT.)
O SSAR STATUS 0 SSAR CHAPTERS SUBMITTED REACTOR & SAFETY SYSTEMS: 9/87 CHAPTERS 4, 5, 6 AND 15 PLANT ARRANGEMENT AND CRITERIA: 3/30 CHAPTERS 1, 2 AND 3 0 NEXT SCHEDULED SSAR CHAPTER SUBMITTAL I&C, AUXILIARY SYSTEM AND QA: 6/30 CHAPTERS 7-9, 11*-14 & 17 0 GE RESPONDED TO FIRST GROUP 0F NRC 4/29
~ ~~
l QUESTIONS 0 NEXT SCHEDULED GROUP OF NRC 7/29 QUESTIONS TO GE
- NUCLEAR ISLAND PORTION OF CHAPTER 11 O
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- O GOOD PROGRESS i
LRB ISSUED 8/87 AND IMPLEMENTED l -
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STAFF /ACRS REVIEW EIGHT ADDITIONAL CHAPTERS BEING READIED FOR 6/88 SUBMITTAL GE RESPONDED TO FIRST GROUP OF NRC QUESTIONS '
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PRESENTED TO ACRS SUBC0l#41TTEE ON THE ADVA! ICED BOILING WATER REACTOR O
JUNE 1, 1988 l
WASHINGTON, D.C.
GE NUCLEAR ENERGY
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O O O 1
- ABWR standard plant design' details with' supporting documentation Reactor building Control building -
l - Nuclear steam supply sys. - Control room l - Primary containment - Plant access control
Change rooms
- Emer. core cooling sys. - Plant supervisors office
- Res. heat removal sys.
Em e r. diesel generators Fuel handling equipment
- Remainder of plant -
safety related 4
interfaces
- Turbine island
- Radwaste facility External p ress u res to expand ABWR sta n d a rd plant to include tu rbin e island and ra d wa ste f acility
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NATIGNAL PERSPECTIVE ON TOTAL PLANT SCOPE FOR CERTIFICATION O
". . . .THE NUCLEAR STEAM SUPPLY SYSTEM OR NUCLEAR ISLAND DESIGN SIMPLY DOES NOT REPRESENT A COMPLETE NUCLEAR THE PLANT. AS SUCH, CERTIFICATION CANNOT BE COMPLETE.
COMMITTEE BELIEVES THAT THE ESSENCE OF THIS PROGRAM MUST BE TO OBTAIN A "RULING" FOR EACH TOTAL PLANT DESIGN IN ORDER TO STANDARDIZE FUTURE PLANTS AND EXPEDITE LICENSING."
FY88 BUDGET REVIEW AND MARKUP BY ENERGY RESEARCH AND PRODUCTION SusCoMMITTEE OF HOUSE SCIENCE AND TECHNOLOGY COMMITTEE. M. LLoYD, SUBCOMMITTEE i
CHAIRMAN. MARCH 1987.
l l "....THE COMMISSION STRONGLY ENCOURAGES THE USE OF CERTIFIED DESIGNS FOR THE ENTIRE PLANT IN ALL FUTURE LICENSE APPLICATIONS...."
".... APPLICATIONS FOR ESSENTIALLY COMPLETE DESIGNS ARE PREFERRED AND WILL BE GIVEN PRIORITY IN ALLOCATION OF RESOURCES TO SUPPORT REVIEW AND APPROVAL."
NUCLEAR POWER PLANT STANDARDIZATION POLICY STATEMENT. (SEPTEMBER 15, 1987; 52FR34884)
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O ABWR SSAR CHAPTER 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS i
l l
PRESENTED TO I O ACRS SUBCOMMITTEE ON THE
- ADVANCED BOILING WATER REACTOR l
l JUNE 1, 1988 j
WASHINGTON, D. C.
i O
GE NUCLEAR ENERGY I
O AGENDA o OVERVIEW o INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY o REACTOR VESSEL I
l i
o COMPONENTS AND SUBSYSTEMS O
O
OVERVIEW i O !
o REACTOR COOLANT SYSTEM COMPONENTS REACTOR VESSEL AND INTERNALS REACTOR INTERNAL PUMPS (10 PUMPS)
MAIN STEAM, SAFETY RELIEF AND FEEDWATER SYSTEMS TO ISOLATION VALVES, WITH AUTOMATIC DEPRESSURIZATION FEATURE REACTOR WATER CLEANUP SYSTEM (1) l HIGH PRESSURE CORE FLOODER SYSTEMS (2) (CHAPTER 6)
RESIDUAL HEAT REMOVAL SYSTEMS (3), WITH LOW PRESSURE l
Q CORE FLOODING FEATURE l
l l
1 O
2 l
OVERVIEW (CONT'D.)
(:) o REACTOR VESSEL AND INTERNALS FEATURES ELIMINATION OF LARGE RECIRCULATION FLOW N0ZZLES BELOW THE CORE BY USING INTERNAL PUMPS IMPROVES CORE FLOODING AND SAFETY PERFORMANCE REACTOR INTERNAL STEAM SEPARATION OF PROVEN BWR6 TYPE LARGE RPV DIAMETER PLACES THE PRESSURE BOUNDARY MATERIAL WELL AWAY FROM THE CORE: THIS REDUCES FLUENCE TO VERY LOW VALUES, THUS GIVING LONG LIFE i
WITH GOOD PROTECTION AGAINST BRITTLE FRACTURE EXTERNAL RECIRCULATION PIPES AND VALVES ELIMINATED INTERNAL PUMPS ARE VARIABLE SPEED TO ADJUST POWER
() WITHOUT ROD MOVEMENT l -
INTERNAL PUMP INERTIA PROVIDES SLOW FLOW COASTDOWN TO l KEEP FUEL WITHIN THERMAL LIMITS FLOW RESTRICT 0RS ARE IN RPV STEAM OUTLET N0ZZLE 1
FLOW RESTRICTORS IN N0ZZLE AND NO EXTERNAL LOOP REDUCE LOSS-OF-COOLANT ACCIDENT LOADS ON REACTOR INTERNALS AND CONTAINMENT l
()
2 l
l
INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY
[])
(SECTION 5.2) o RCPB COMPLIES WITH 10CFR50, SECTION 55A, CODE REQUIREMENTS INCLUDING CODE CASE APPROVAL OF CL. 1, 2 & 3 COMPONENTS.
o OVERPRESSURE PROTECTION CONFORMS TO 10CFR50, APP. A, GDC 15 o AUTOMATIC DEPRESSURIZATION SYSTEM INCLUDED WITH 3 LOW PRESSURE FLOODER SYSTEMS AND 3 HIGH PRESSURE FLOODER/ FEED SYSTEMS (RCIC + 2 HPCF's) o SAFETY RELIEF VALVES DISCHARGE TO PRESSURE SUPPRESSION P0OL AND LIMIT PRESSURE TO 110% DESIGN PRESSURE PER ASME l CODE SECT. III FOR MSIV CLOSURE WITH HIGH FLUX SCRAM.
O o SRV's OPEN ON PRESSURE ACTING AGAINST SPRING FORCE (SAFETY MODE), AND BY PNEUMATIC ACTUATOR (RELIEF MODE).
l o RELIEF MODE OF OPERATION CV.S'dIONS CLOSING FORCE AND MAINTAINS RE-CLOSING SEQUENCE.
o THE ADS MAKES USE OF 8 0F 18 SRV's OPERATED BY THE PNEUMATIC ACTUATORS.
C:)
l l
3 l
i 1
'"""'" ' "C"" (C " -)
p O (SECTION 5.2)
{
HAVE SUCCESSFUL OPERATING EXPERIENCE i -
BE FULLY QUALIFIED IGSCC RESISTANT INCLUDE KNOWN METALLURGICAL IMPROVEMENTS o ENVIRONMENTAL CONTROLS INTEGRATED WITH MATERIAL SELECTION O -
EPRI BWR OWNERS WATER CHEMISTRY GUIDELINES HYDROGEN WATER CHEMISTRY TO PROVIDE ADDITIONAL MARGIN AGAINST SCC UTILIZE PROVEN MATERIALS AND PROCESSES O
4 l
l Q INTEGRITY OF RCPB (CONT'D.)
(SECTION 5.2)
MATERIALS QUALIFIED FOR GE ABWR e PROVEN RELIABLE THROUGH OPERATIONAL SERVICE l
l 304L/316L STAINLESS STEEL LOW CARBON STAINLESS STEEL CASTINGS XM-19 STAINLESS STEEL CASTINGS
- CARBON STEEL l
A533/A508 LOW ALLOY STEEL UNCREVICED ALLOY 600 Qo QUALIFIED THROUGH EXTENSIVE DEVELOPMENT TESTING l -
304/316 NUCLEAR GRADE STAINLESS STEEL HIGH TOUGHNESS CARBON STEEL l
RADIATION RESISTANT LOW ALLOY STEELS CONTROLLED COMPOSITION ALLOY 600 AND WELD METALS SPECIAL HEAT TREATED X-750 HIGH PURITY STAINLESS STEEL FOR HIGH FLUENCE COMPONENTS o ADDITIONAL DESIGN CONTROLS CREVICE GEOMETRIES AVOIDED COLD WORK / FABRICATION CONTROLS COMPOSITION OF WELD METALS 5
l l
4 REACTOR VESSEL
(]) (SECTION 5.3) o INCREASED DIAMETER (27 INCHES) FOR PUMP REMOVAL FURTHER REDUCES NEUTRON FLUENCE 17 6 X 10 FOR 60 YEARS 0
MAXIMUM FINAL RT NDT AFTER 60 YEARS 50 F FRACTURE MECHANICS EVALUATION WITH ASSUMED FLAWS SHOW NO CRACK PROPAEATION o IN-SERVICE EXAMINATION NO WELDS IN CORE BELTLINE NO WELDS IN BOTTOM HEAD SECTION CONTAINING CRD PENETRATIONS
(]) -
CIRCLE SEAM BETWEEN CRD's AND RIP INSPECTABLE WITH REMOTELY OPERATED UT POSITIONER IMPROVED ACCESS TO REACTOR VESSEL INSIDE l (:)
6
i l COMPONENTS (1 (SECTION 5.4) l REACTOR INTERNAL PUMPS -- KEY FEATURES o WET MOTOR, SEAL-LESS' DESIGN WITH CASING WELDED TO THE RPV o SOLID STATE ADJUSTABLE FREQUENCY SPEED CONTROL o CONTINUOUS PURGE WITH CLEAN WATER FROM CONDENSATE STORAGE TANK l
o MOTOR COOLED BY REACTOR BUILDING CLOSED COOLING WATER
($) SYSTEM VIA CLOSED COOLING LOOP DN THE RIP (1 HX PER PUMP) o IMPELLERS AND MOTORS REMOVABLE WITHOUT REACTOR DRAINING o BACK SEATING SHAFT AND BLOW OUT RESTRAINT HANGERS PROVIDE REDUNDANT LOCA PREVENTION l
l l
7 l
- REACTORINTERNALPU$P_
O ,_
SHROUD SUPPORT
( '
= PUMP IMPELLER
'?
I I
I : DIFFUSER
.q p a h I
- REACTOR VESSEL h PURGE WATER INLET l
~ j?
l COOLING WATER
- *<n~,
q g ,
OUTLET jF -
N, MOTOR CASING O - .,
i
= PUMP SHAFT l
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ROTOR SHAFT STATOR l .
7 : 1 F 7 SHAFT COUPLING
- I .
- STUD
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THRUST BEARING Yb I g x w3:w:
p y.. .
r COOLING WATER Q,&' .
,]g,y INLET 8
l l
l
- O O O j EVALUATION OF REACTOR INTERNAL PUMPS i
TYPICAL BWR ABWR COMENTS DRYWELL ARRANGE- RECIRCULATION PUMP, CONSIDERABLE SIMPLI- THIS IS A MAJOR j MENT VALVES, PIPING AND FIED. PERMITS UPPER RIP ADVANTAGE MOTOR (INCLUDING AND LOWER DRYWELL REMOVAL REQUIRF- ARRANGEMENT RATION-MENTS) HAVE A MAJOR ALIZATION ,
INFLUENCE ON DRY-WELL ARR/.NGEMENT SAFETY (ACCIDENT) LARGE N0ZZLES ARE NO LARGE N0ZZLES RIP HAS SUPERIOR PERFORMANCE REQUIRED BELOW THE BELOW THE CORE. LOCAL PERFORMANCE.
TOP OF THE CORE. MODEST ECCS PERMITS REDUCED THIS RESULT IN CAPACITY WILL' ECCS CAPACITY.
RAPID LOCA FLUID KEEP THE CORE LOSSES AND CORE COVERED FOR ALL UNC0VERY/ FUEL HEAT LOCA CONDITIONS UP DURING THE ACCIDENT 9
^ '
O EVALUATION OF_ REACTOR INTERNAL PUMPS u COPMENTS o
TYPICAL BWR ABWR LARGE PUMP / MOTOR SPECIALIZED UNDER- RIP MAINTENANCE PUMP AND MOTOR UNITS IN DRYWELL: VESSEL RIP MOTOR IS EASIER, LESS MAINTENANCE HIGH RADIATION REMOVAL DEVICES OPERATOR EXPOSURE FIELD, DIFFICULT TO FACILITATE MAIN- NOT ON CRITICAL TO HANDLE. TENANCE (SPECIALIZED PATH f TOOL, CARTS, AND HOIST). PROVEN BY f T&D PROGRAMS AND EUROPEAN EXPERIENCE.
MOTORS SERVICE OUT-SIDE CONTAINMENT.
i IN-SERVICE EXTENSIVE EXTERNAL NO CODE REQUIRED SIGNIFICANT PIPING REQUIRING ISI ON THE RIP REDUCTION IN INSPECTION (ISI)
ISI. SUBSTANTIAL CASINGS OR N0ZZLE OPERATOR EXPOSURE.
PERSONAL EXPOSURE. WELDS.
WITH ONE PUMP OUT 100% POWER CAN BE IMPROVED PLANT PLANT AVAIL-ABILITY AND OF SERVICE, OUTPUT MAINTAINED CONTIN- AVAILABILITY. ,
MARGINS LIMITED TO 60-70%. UOUSLY WITH ONE PUMP OUT OF SERVICE.
111% MAXIMUM CORE FLOW CAPABILITY WITH ALL PUMPS. SOLID STATE CONTROLLERS.
10 l
~
l
O O O EVALUATION OF REACTOR INTERNAL PUMPS TYPICAL BWR ABWR C0fMENT_5 ANNULUS THES LOCA-RELATED RIP BLOWOUT RIP RESULTS IN L PRESSURIZATION ISSUES HAVE BEEN RESTRAINTS PROVIDED. DESIGN SIMPLIFI-l ADDRESSED IN THE NO PIPE WHIP OR CATION AND l DESIGNS. RESOLU- ANNULUS PRESSURI- REDUCED CONTAIN-
! TION INVOLVES ZATION ISSUES MENT CONGESTION.
DIFFICULT AND ASSOCIATED WITH j EXPENSIVE ITEMS LARGE PIPE RUPTURES.
l (RESTRAINTS, FLOW DIVERTERS) l PUMP SEALS MECHANICAL SHAFT SEAL-LESS WET RIP WILL HAVE
} SEAL; SOMETIMES MOTOR WITH LESS MAINTENANCE, j WITH SEAL PURGE. PURGE. LEAKAGE AND CON-l THESE SEALS HAVE TAMINATION.
l BEEN A SIGNIFICANT ENHANCED PLANT
! SOURCE OF FORCED AVAILABILITY.
DUTAGES.
i i
j j 11
l REACTOR PRESSURE COMPACT JET PUMP USE OF RIP REQUIRES CURRENT JET PUMP l VESSEL (RPV) SIZE DESIGN PERMITS LARGER DIAMETER DESIGN IS MORE MINIMUM RPV (APPROX. 27 INCHES) THAN ADEQUATE IN
- DIAMETER FOR PUMP IMPELLER TERMS OF VESSEL l REMOVAL. HOWEVER, WALL NEUTRON j THIS GIVES THE FLUENCE AND RPV
! BENEFITS OF: PRESSURE RATE.
! - REDUCED NEUTRON HOWEVER, THE RIP
{ FLUX AT THE VESSEL DESIGN GIVES A BELTLINE. SIDE BENEFIT OF ll - REDUCED REACTOR ADDED MARGIN IN PRESSURE RATES THESE AREAS.
- DURING TRANSIENTS.
! THIS RESULTS IN j LESS S/R VALVE l LIFTS.
I l
PLANT HEAT RATE: 14 MW 10MW RIP PLANT HAS PUMPING POWER LOWER HOUSE LOAD.
I i
i
O O O EVALUATION OF REACTOR INTERNAL PUMPS TYPICAL BWR ABWR COMMENTS j CORE FLQW M-G $ETS OR FLOW SOLID STATE, RIP HAS IMPROVED l CONTROL CONTROL VALVES ADJUSTABLE, OPERABILITY AND l PROVIDE FULLY FREQUENCY GIVES MAINTENANCE, MORE ADEQUATE CORE STEPLESS, GANGED PRECISE FLOW FLOW CONTROL SPEED CONTROL CONTROL.
i FROM 30% TO 112%
l RATED FLOW. MORE j RAPID CORE FLOW j RESPONSE (1% vs.
l 1/2% SEC.) N0 i
DEAD BANDS OR DELAY TIMES. SUPPORTS li PLANT AUTOMATION GOALS.
13
I l
O REACTOR INTERNAL PUMP OPERATING EXPERIENCE AND T7D BASE OPERATING EXPERIENCE o NEARLY 100 PUMPS IN SERVICE IN EUROPEAN BWRs o APPROACHING 600 PUMP-YEARS OF EXPERIENCE o ONLY ONE INCIDENT OF A MAJOR FORCE OUTAGE DUE TO PUMP-RELATED PROBLEMS (BRUNSBUTTEL IN 1977)
_T&D BASE o EXTENSIVE EUROPEAN TESTING OVER THE YEARS BY
[]) SEVERAL PUMP VENDORS AND NSSS SUPPLIERS o SEPARATE TESTING BY HITACHI AND TOSHIBA o JAPANESE GOVERNMENT AND UTILITY TESTING PROVEN TECHNOLOGY 4
O 14
- l8 O
f ABWR SSAR CHAPTER 6 ENGINEERED SAFETY ~ FEATURES PRESENTED TO O ACRS SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR JUNE 1, 1988 WASHINGTON, D.C.
O .
GE NUCLEAR ENERGY
l i 0VERVIEW
! ()
l o ENGINEERED SAFETY FEATURE MATERIALS l i
o* CONTAINMENT SYSTEMS o* EMERGENCY CORE COOLING SYSTEMS o HABITABILITY SYSTEMS SECTION 6.4 PRESENTS CRITERIA SECTION 9.4.1 PROVIDES DETAILS (LATER
() SUBMITTAL) o* STANDBY GAS TREATMENT SYSTEM (SGTS) o IN-SERVICE INSPECTION OF CLASS 2 AND 3 COMP 0NENTS o NITR0 GEN GAS SUPPLY SYSTEM
- PRESENTATIONS TO F0LLOW
()
SIPEl:CDS:FL880601: Jow
STANDBY GAS TREATMENT SYSTEM O
o ADVANCE FILTER TRAIN DESIGN /0PERATION CONFIGURATION OF FILTER HOUSING AND FLOW PATTERN VIRTUALLY ELIMINATES ANY UNTREATED BYPASS OF THE FILTERS ELIMINATION OF INADVERTENT WATER SPRAY OF THE CHARC0AL BY NOT NORMALLY CONNECTING A WATER SOURCE TO THE DELUGE PIPING WATER HOSE AND CONNECTIONS NEARBY PERIODIC TESTING 0F THE FILTERS TO ENSURE l CHARC0AL ABSORBER AND FILTER EFFECTIVENESS ELIMINATES AGING EFFECTS o SIGNIFICANTLY IMPROVED LONG-TERM PERFORMANCE CAPABILITY AND RELIABILITY ONE FILTER SATISFIES STGS FUNCTION O
SIPEl:CDS:FL880601: JDw
O O O D = DEMISTER l@ (FROMPRIMARYCGCS CONTAINMENT) E = ELECTRIC PROCESS HEATER S P = PREFILTER
$ H C
= HEPA FILTER _4
,_ = CHARCOAL ADSORBER $
~
1 I ECONDARY = SPACE HEATER CONTAINMENT d $ h= BACKDRAFT GRAVITY DAMPER E
TO STACK
~~
1 kh J k f(
D E 7_
l
@ G
- P H C H -
~
l I
FILTER TRAIN $
D E DRYER TRAIN 8 F J9fie t 5-b >
w Figure 65-1 SCHEMATIC DIAGRAM OF STANDBY GAS TREATMENT SYSTEM l . - _ _
I' q O
CONTAINMENT SYSTEM (SECTION 6.2)
PRESENTED TO ACRS SUBCOMMITTEE ON THE A VANCED BOILING WATER REACTOR O
JUNE 1, 1988 WASHINGTON, D.C.
O GE NUCLEAR ENERGY l
ABWR CONTAINMENT
()
PRIMARY CONTAINMENT STRUCTURE o REINFORCED CONCRETE CYLINDER 95 FT ID X 96.8 FT HIGH o STEEL LINED o STRUCTURALLY INTEGRATED WITH SURROUNDING REACTOR BUILDING & UPPER POOLS o SEISMIC DESIGN 0.3G
() o DESIGN LOADS i DEAD LOAD + THERMAL LOCA DYNAMIC (C0 & CH) l SRV P00L SWELL o DESIGN PRESSURE 45 PSIG & (-) 2 PSID SECONDARY CONTAINMENT (REACTOR BUILDING) SURROUNDS PRIMARY CONTAINMENT C)
CCE3: GEW:FL880601: JDW l . _ - . _ - . - _ . - - -
..g+
a ABWR CONTAINMENT O CONTAINMENT CHAMBERS Cu FT o UPPER DRYWELL 95 FT OD X 29.5 FT HIGH 223,000 o LOWER DRYWELL 34.8 FT OD X 38 FT HIGH 36.00_Q TOTAL 259,600 o SUPPRESSION CHAMBER 95 FT OD X 46 FT ID AIRSPACE 40 FT HIGH 210,500
~
POOL 23 FT DEEP 126,400 0
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HORIZONTAL VENTS
^8" O CONTAINMENT CONFIGURATION
q ABWR CONTAINMENT O
PRESSURE SUPPRESSION o HORIZONTAL VENTS CONNECT DRYWELL TO SUPPRESSION CHAMBER INTEGRAL WITH RPV PEDESTAL AS AN EXTENSION
! 0F THE DRYWELL CONNECTING VENTS MARK III CONFIGURATION & DATA BASE IAMETER O -
SPACING SUBMERGENCE TEPC0 SPONSORED FULL & SCALE TESTS EVALUATED BACK PRESSURE EFFECT ON C0 & CH LOADS
\ .
1 0
CCE3: GEW FL880601:aow
ABWR CONTAINMENT
() LOCA RESPONSE PEAK VALUES LOCATION CALCULATED DESIGH DRYWELL PRESSURE 39 PSIG(1,2) 45 PSIG DRYWELL TEMPERATURE 33 0F C} 3400F WETWELL PRESSURE 26 PSIG(1,2) 45 PSIG WETWELL TEMPERATURE 2070 F(l) 2190F l
()
l l
FLOW LIMITING ORIFICE LOCATED IN RPV N0ZZLE CALCULATED VALUES ENVELOPE MAXIMUM OF STEAM OR FEEDWATER LINE BREAKS O
(1) FEEDWATER BREAK (2) SHORT TERM C) 7
@ h@5. i f7 - -
CCE3: GEW:FL880601:anw
ABWR CORTAINMENT VACUUM BREAKER SYSTEM O
o CONNECT SUPPRESSION CHAMBER AIRSPACE WITH
- LOWER DRYWELL i
i
! o PENETRATIONS SEPARATE FROM VENT SYSTEM i
! o EIGHT 20" VBs PERFORMANCE o REMAIN CLOSED DURING CHUGGING & POOL SWELL O o REMAIN CLOSED ON P0OL SWELL o OPEN ON DRYWELL SPRAY ACTUATION OF ECCS FLOW OUT BREAK o SEVEN VALVES PROVIDE A /K = 8.3 FT2 o DW TO SC DIFFERENTIAL PRESSURE ,
DESIGN (-) 2 PSID CALCULATED (-) 1.5 PSID O
CCE3: GEW:FL880601: Jow
4
() ABWR CONTAINMENT -
- BYPASS LEAKAGE - DRYWELL TO SUPPRESSION LHAMBER o DIAPHRAGM FLOOR RIGIDLY CONNECTED TO CONTAINMENT o DIAPHRAGM FLOOR STEEL LINED o RPV PEDESTAL FABRICATED STEEL STRUCTURE FILLED WITH CONCRETE ,
()
STEEL LINED INTERFACE BETWEEN DW & SC SUPPRESSION CHAMBER SPRAYS OPERATE FROM RHR SYSTEMS B&C SPRAY BYPASS CAPACITY A V K = .0E 52 9
()
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ABWR CONTAINHEHI (2) SECONDARY CONTAINMENT o SURROUNDS PRIMARY CONTAINMENT o AIRLOCKS AND SEALED PENETRATIONS o FLUID SYSTEMS TO CLEAN AREA HAVE WATER SEALS OR ISOLATION VALVES o OPERATES AT NEGATIVE PRESSURE RELATIVE TO PRIMARY CONTAINMENT & REACTOR BUILDING CLEAN ZONES (b o MAXIMUM IN LEAKAGE 50% 0F SECONDARY l CONTAINMENT VOLUME / DAY o COMPARTMENTS VENTED FOR HIGH ENERGY LINE
- BREAKS 1
o AIR FLOW FROM TOWARD MOST HIGHLY CONTAMINATED AREAS 1
l 0 EXHAUST MONITORED & CAN BE ROUTED THROUGH '
()
l I CCE3: GEW:FL880601: JDW l
l 1 .
MlD '
- u. . l 1
[
l O !
EMERGENCY CORE COOLING SYSTEMS (SECTION 6.3)
PRESENTED TO ACRS SUBCOMMITTEE ON THE O
ADVANCED BOILING WATER REACTOR l
l JUNE 1, 1988 WASHINGTON, D.C.
O
_ _GE NUCLEAR ENERGY l
1
C
(
ECCS KEY PERFORMANCE FEATURES TYPICAL TYPICAL B W R/4 BWR/S BWR/6 ABWR
\ HPCS HPCF LPCI~
HPCI LPCS LPCS LPCI LPCI LPCI LPCI HPCF RCIC LPCI LPCI LPCS LPCI LPCI LPCI ADS ADS ADS i i l HIGH PRESSURE $ 5000 1.900 2800 CAPACITY, GPM O tow PRESSURE **
CAPACITY, GPM 42000 2,000 1,000 NO. OF LARGE PIPES 12 12 0 BELOW CORE PEAK CLAD TEMP 1600 1100 NO UNCOVERY (APP K), 'F N 2 CAPABILITY NONE ALL BUT ALL BUT HPCS LPCI (A)
BREAK BREAK
- @ 1100 PSI
@ 100 PSI -
0 1
, , , , , . ~ - - - . -
l 1
()
CORE SPRAY FUNCTION ELIMINATION o OVERHEAD CORE SPRAY SPARGER REPLACED BY INSIDE SHROUD FLOODER SPARGER o CONSISTENT WITH EPRI ALWR REQUIREMENTS, BASED ON US 'JTILITY GUIDANCE o SIMPLIFICATION OF MAINTENANCE AND SURVEILLANCE o DESIGN BASIS ACCIDENT ANALYSES ESSENTIALLY O UNCHANGED l
()
SIPEl:CDS FL880601: JDW
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ABWR EMERGENCY CORE COOLING SYSTEM High Pressure i,
1 l
{KWoMozoMMfMkMbMMjl
, CONTAINMENT $
l
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]
.j REACTOR .,
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j
- d N . 1 HPCF p [ ~~ g?^gTUR9#8E # RCIC e 2 DMSIONS
- 1 DMSION e 800-3200 s A' L
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l O O O l ABWR EMERGENCY CORE COOLING SYSTEMS
) Low Pressure I
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ACCIDENT PERFORMANCE l
j LOCAs i
! o NO DESIGN BASIS ACCIDENT LEADS TO CORE UNC0VERY LIMITING EVENT IS HPCF BREAK s
o CONSERVATIVELY TRIP ALL RECIRCULATION PUMPS i
AT T=0 o RESULTS i
(S) -
STEAMLINE BREAK < 10400F LIQUID LINE BREAK < 10100F 95% B0llNDING CASE < 11100F
()
SIPE1 CDS:FL880601: Jow
w-y&
l O l ABWR SSAR CHAPTER 15 ACCIDENT ANALYSIS PRESENTED TO ACRS SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR JUNE 1, 1988 WASHINGTON, D.C.
O GE NUCLEAR ENERGY
-,,,,,.-,,-,_..,..,,,,a ,- .... . . - _ .- . , , , ,- - - , - , - , - --,, ,, - - - - , . , - , - . , - , -
w
-. ---,--e e .,, - - - - -- , , , - - ---,_ , ,- -
\
OVERVIEW OF CHAPTER 15
()
o COVERS TRANSIENTS AND ACCIDENTS o EVENTS PACKAGED i -
DECREASE IN COOLANT TEMPERATURE INCREASE IN PRESSURE DECREASE IN FLOW RATE
) REACTIVITY ANOMALIES INCREASE IN COOLANT INVENTORY DECREASE IN COOLANT INVENTORY l
l ANTICIPATED TRANSIENTS WITHOUT SCRAM o NUCLEAR SAFETY OPERATIONAL ANALYSIS PROVIDED .
()
SIPE1: CDS:FL880601: Jow
ABWR FUEL fs '
\J o BWR STANDARD FUEL DESIGNS USED SAR ANALYSIS BASED ON 8X8 BARRIER FUEL CONTROL CELL CORE FUEL MANAGEMENT o CONTROL ROD PITCH SLIGHTLY INCREASED REDUCED VOID COEFFICIENT o STEADY-STATE PERFORMANCE RESULTS (EQUILIBRIUM CYCLE):
LIMIT MINIMUM CRITICAL POWER RATIO 2 1.44 1.16
- MAXIMUM LINEAR HEAT, KW/FT $ 11.1 14.4 COLD SHUTDOWN K Err 4 0.96 0.99 i
SUBSTANTIAL CORE DESIGN MARGINS O
SIPEl:CDS:FL880601: Jow
ABM z wiooAn Standard Plant nEv A I
- - i NOTE: LOADING PATTERN IS SHOWN FOR '
~
OUARTE R CORE ONLY. ROTATIONAL --
SYMMETRY TMLIES. 4 4 4 2 2 1 2 j 4 2 2 1 1 2 2 3 4 4 4 1 2 1 1 2 1 2 4 I
4 2 2 1 2 1 2 3 1 2 1 5 4 2 1 1 2 1 2 2 1 2 3 2 6 4 2 2 1 2 1 2 3 1 2 3 1 3 7 4 2 1 1 2 1 1 2 1 2 3 1 2 1 8 4 2 1 2 1 3 3 1 2 3 1 2 1 3 9 4 1 2 1 1 3 3 1 3 1 2 2 1 3 10 4 1 2 1 2 2 1 1 2 1 2 3 1 2 1 11 4 2 2 1 2 3 1 2 3 1 2l 1 1 2 3 2 12 4
l 2 1 l2 2 1 2 3 1 2 1 3 3 2 1 3 13 4 1 1 3 1 2 3 1 2 3 1 3 3 1 2 1 14 4 2 1 2 1 2 3 1 2 2 1 2 2 1 2 3 2 15 4 1 2 1 2 3 1 2 1 1 : 3 1 2 3 I 1 2 16
_1 s_
4 2 1 l2 3 1 2 2 3 l
3 2 l1 2 l
3 1 2 3 17 Ja-1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 BUNDLE TYPE ENRICHMENT NUMBER OF BUNDLES 1 HIGH ENRICHMENT,3.18 wt % 303 2 MEDIUM ENRICHMENT,2.19 wt % 324 3 LOW ENRICHMENT,1.23 wt % 148 4 N ATUR AL UR ANIUM,0.71 wt % 92 O
V 87 259 01 Figure A.3-1 INITIAL CORE LOADING MAP 4M
TRANSIENTS-KEY PERFORMANCE FEATURES O OPERATING BWR ABR8 COMMENT STARTUP NEUTRON IRM/ RANGE WRNM/ PERIOD MONITORING SWITCHES BASED PROTECTION ELIMINATES Roo WITHDRAWAL ERROR POWER RANGE Roo Roo BLOCK ADVANCED Roo CONTROL MONITOR BLOCK MONITOR j CONTROL SYSTEMS 1-CHANNEL 3-CHANNEL MAKES CONTROL ANALOG DIGITAL SYSTEM SINGLE FAILURE PROOF Vozo COEFFICIENT, -14 -11.6 t/%
O LEVEL RATE, -7.2 -5.7 j GENTLER IN/SEC. TRANSIENTS PRESSURE RATE, 160 140 s PSI /SEC ABWR SIMPLIFIES AND REDUCES SEVERITY OF TRANSIENTS O
SIPEl:CoS:FLB80601: JoW
A TRANSIENT PERFORMANCE O
OVERPRESSURE PEAK PRIMARY PRESSURE P_1LG o REQUIREMENT f 1375 o CLOSURE OF A TURBINE CV 1113 O o GENERATOR LOAD REJECTION / 1220 TURBINE TRIP o MSIV CLOSURE 1242 o MSIV CLOSURE - BACKUP SCRAM 1274 (ASME EVENT) 6 SIGNIFICANT MARGIN TO LIMIT O
SIPE1 CDS:FL880601:aow
TRANSIENT PERFORMANCE l
() l CRITICAL POWER 4_CP_R o DECREASE IN COOLANT TEMP RUN0UT OF FW PUMP 0.06 o INCREASE IN REACTOR PRESSURE CLOSURE OF TURBINE CV 0.09 o DECREASE IN COOLANT FLOW RATE TRIP OF 3 REACTOR INTERNAL PUMPS 0.04 O o REACTIVITY ANOMALIES R0D WITHDRAWAL ERROR ELIMINATED o INCREASE IN COOLANT INVENTORY INADVERTENT HPCF INITIATION NO CONSEQUENCE o STATISTICAL EVALUATION 0.07 l
l l OPERATING LIMIT CPR SET AT 1.16 C) .
i .
SIPEl:CDS:FL880601: Jow -
1
. - . . - - - - - . - . - . . _ . . - _ . - . . . . . . _ - - . - _ . - ~ - . . - - . - _ _ . _ _ . - . - - . . . . - . . - . - . - . .
lO ACCIDENT PERFORMANCE i OTHER EVENTS
O (MULTIPLE FAILURES) <100 CAL /GM f
R0D EJECTION (MULTIPLE FAILURES) < 70 CAL /GM o REACTOR INTERNAL PUMP SEIZURE NO CONSEQUENCES l
O SIPEl:CoS:FL880601: Jow
RADIOLOGICAL EVALUATIONS
SUMMARY
() DOSE, REM
- THYROID WHOLE B02X o REACTIVITY ACCIDENTS BOUNDING ANALYSIS, BASED ON 2.4 .07 280 CAL /GM PEAK PELLET, EVEN THOUGN ANALYSIS SHOWS
<100 CAL /GM 0 LOCAS INSTRUMENT LINE BREAK OUTSIDE NEGL. NEGL.
STEAMLINE BREAK OUTSIDE 0.6 0.01 FEEDWATER LINE BREAK OUTSIDE 0.4 NEGL.
DBA**
O CONTROL ROOM (30-DAY) 0.03 2 0 SITE BOUNDARY (300M, 2 HRS) 1.5 0.6 0 LOW POPULATION ZONE 22 12 l (800M, 30-DAY) 0 FUEL HANDLING FUEL BUNDLE DROP 0.4 0.4 CASK DROP 0.01 0.001
- AT 1/2 MILE WHERE APPROPRIATE
- DEVIATIONS FROM RG 1.3; FOLLOWS NUREG 1169 O
SIPE1lCDS:FL880601:aDW
,1
- - , n,,--._, --
i ATWS COMPLIANCE
()
o FMCRD HAS TWO DIVERSE INSERTION MODES l HYDRAULIC SCRAM ELECTRIC RUN-IN (APPR0X. 3 MIN)
ADDRESSES PREVIOUS COMMON-MODE FAILURE CONCERNS o ABWR HAS RECIRC PUMP TRIP (RPT) o SLCS NOT NEEDED FOR ATWS, BUT MANUAL CAPABILITY AVAILABLE l () o NO FORMAL ANALYSIS SUBMITTED BUT STUDY RESULTS SHOW PEAK PRESSURE 1340 PSIG 8% RODS IN TRANSITION BOILING PEAK P00L TEMPERATURE 1600F WELL WITHIN PREVIOUS INDUSTRY /NRC STUDIES WHICH LED TO 10CFR50.62
()
SIPE1lCDS:FL880601:aDw
NUCLEAR SAFETY OPERATIONAL ANALYSIS
() (NSOA) o PURPOSES:
TO IDENTIFY SAFETY FUNCTIONS TO DEMONSTRATE CONPLIANCE TO SINGLE FAILURE CRITERIA o PROCESS:
DEFINE EVENTS
() -
DEFINE ACCEPTANCE CRITERIA DEFINE SUCCESS CRITERIA INCLUDE ALL PLANT STATES INCLUDE SUPPORT SYSTEMS o 56 INDIVIDUAL EVENTS STUDIED
()
SIPEl:CDS:FL880601:aDw '
ABM z w iooxa Standard Plant arv A UNACCEPTABLE RESULTS O NUCLE AR S AF ETY PL ANT SYSTEMS AS DESIGNED NUCLEAR SAF ET Y OPERATIONAL OEsiGN CRITERI A AND INST ALLE O CRITE RI A O
NORMAL FRE UENCY INFREQUENT LIMITING SP ECI AL OP E R ATION S INCIDENTS INCIDENTS F AULTS EVENTS BWR ERAT AL OPE R ATING 4gg(
STATES I
RULES FOR spENTIFICATION OF SAF ETV ACTIONS EVEN1 ESSENTI AL TO AVOIDING UN ACCEPTABLE ANALYSl$ RESULTS IE ACH STATEl j REDUNDANCY l IDENTIFIC ATION OF SYSTEMS AND (SACF LIMITS ESSENTI AL FOR ACHIEVING
.5AF ETY ACTIONS IE ACM STATEl I
IDENTIFIC ATION OF INTR ASYSTEM ACTIONS WwiCM MUST BE REQUIRED OR RESTRICTED IE ACM SYSTEM)
I IDENTIFIC ATION OF MINIMUV SYSTEM H ARDWARE CONDITIONS TO ACCOWLISH ACTIONS IE ACM SYSTEVI O i IDENTIFICATION OF H ARDWARE CONDITIONS TO $ ATISF Y REDUNDANCY REQUIREMENTS E ACM SYSTEM (OPER ATION AL NUCLE A9 SAFETY REQUIREVENTS LIMITING CONDITIONS FOR OPER ATIONI l
IDENTIFY SURVEILLANCE "
TEST F RE QUENCIES i j . AV AIL ABILITY CONSIDE R ATIONS l
i IDt.4TIF Y ALLOW ABLE REPAIR YlMES WHEN SYSTEM INOPER ABLE ILIMITING I CONDITION FOR OPER ATIONI l l ACTION TO BE T AKE N I IF OPER ATION AL REOuiREMENTS NDT SATISFIED SIMPLIF IC ATION r--- -------- -------- - - - - - - - -
7 l SYSTEM LEVEL I8 L SYSTEM LEVEL CH C l OV ALIT ATiv E QU ALIT ATIV E g, p q l OU VE Y EM PL AUT S AF ETY g PLANT DESIGN B ASl5 l CON FIRM ATION DEslGN B Asis l CONFIRM ATION TOTAL SAF ETY l R E WIR EM E NT ,,j l L - .. . _ _ .
B7 258 :6 Q Figure 15A.2-1 BLOCK DIAGRAM OF METHOD USED TO DERIVE NUCLEAR SAFETY OPERATIONAL REQUIREMENTS SYSTEM LEVEL QUALITATIVE DESIGN l BASIS CONFlRMATION AUDITS AND TECHNICAL SPECIFICATIONS P
1SA19 I
.%va rLT
\ 2-ACRS INQUIRY ON HYDROGEN DETONATION AND DAMAGE TO SAFETY RELIEF VALVES E
DAMAGE TO SEVERAL SAFETY RELIEF VALVES (SRVs) RESULTING FROM HYDROGEN DETONATION IN OVERSEAS BWR PROBABLE CAUSE FAST COMPRESSION OF HYDROGEN AND OXYGEN (GENERATED BY RADIOLYTIC DECOMPOSITION OF WATER MIGRATED FROM REACTOR CORE TO MAIN STEAM LINES) BY ACTIVATION OF PILOT VALVE LINE LEADING TO SRV INTERNALS O
CORRECTIVE ACTION PREVENT HIGH CONCENTRATION OF HYDROGEN AND OXYGEN OR ACTIVATE VALVE WITHOUT COMPRESSION OF MIXTURE ABWR STATUS DIRECT ACTING SRV WITH DECOMPRESSION O
W O O
- . . O .
ABWR Licensing Scene ,
.; .- Licensing Schedules
- .~,
.;- +,
' .i E- 1993 1995 1997 4 1987 1989 1991 I
- h h2 I l l l l l l l 1 I I I
.4 4f. , t .
Er .E? '
_ Receipt of EP
'Q $]: .
'4 WO Application for CP and K-6 Construction
- < L Begins
%: ^~ 9-
~ ~ '
Establishment K-6 Commercial Permit (EP) Operation A A A -
A
.... MITI i a g.y :
T hO5 k A K-7 Construction K-7 Commercial l
Begins Operation
_ Ucensing Review Basis l' ~' lasued ACRS Letter NRC Certification
^^ ^ ^ ^
NRC
> 1 l .
_ Final Design Approval lasued I
Application for Certification A
I I I I I I I I I I i I I 1988 1990 1992 1994 1996 1938 l
W O
i l
ABWR SSAR CHAPTER 4 REACTOR DESIGN l -
1 l
l PRESENTED TO ACRS SUBCOMMITTEE ON THE O ADVANCED BOILING WATER REACTOR l
JUNE 1, 1988 WASHINGTON, D.C.
l l
O GE NUCLEAR ENERGY l
ACRS SUBCOMMITTEE ON THE ADVANCED BOILING WATER REACTOR O JUNE 1, 1988 CHAPTER 4 -
REACTOR OVERVIEW CONSISTENT WITH THE SRP CHAPTER 4 ADDRESSES:
O REACTOR ASSEMBLY (RPV & W ERN RS)
FUEL SYSTEM DESIGN REACTOR THERMAL-HYDRAULIC DESIGN REACTOR MATERIALS REACTIVITY CONTROL SYSTEMS O
~
REACTOR VESSEL AND INTERNALS O DESIGN OBJECTIVES 0 DEFINE A REACTOR SYSTEM WHICH FULLY CONTRIBUTES TO ALL MAJOR ABWR OBJECTIVES
- DESIGN SIMPLIFICATION
- ADDITIONAL MARGINS l
l REDUCED MAINTENANCE COSTS LOWER OCCUPATIONAL EXPOSURE O -
CAPACITY FACTOR / AVAILABILITY IMPROVEMENTS l
ENHANCED REACTOR SAFETY 0 ADDRESS ALL BWR MATERIALS CONCERNS 0 USE ESTABLISHED TECHNOLOGY OR TEST BEFORE USE l
O l
1
.. - ~. . _ . . - - _ - .
REACTOR KEY DESIGN FEATURES 0 REACTOR PRESSURE VESSEL
(])
REDUCED RADIUS CLOSURE HEAD AND FLANGES RIP PENETRATIONS MACHINED IN BOTTOM HEAD KNUCKLE FORGING STEAM FLOW RESTRICTOR IN N0ZZLE-ALSO USED FOR FLOW MEASUREMENT RPV VENT AND HEAD SPRAY N0ZZLE IN TOP HEAD 0 REACTOR INTERNALS .
BWR/6 TYPE STEAM DRYER AND STEAM SEPARATORS l
BORON CAF. BIDE CONTROL ROD WITHOUT VELOCITY LIMITER (1 FOOT SHORTER)
({}
BWR/6 TYPE CORE STRUCTURE WITH TOP GUIDE MACHINED FROM THICK PLATE AND CORE PLATE l STIFFENERS IN TWO DIRECTIONS SEPARTE LOW PRESSURE FLOOD SPARGER IN VESSEL AND HIGH PRESSURE CORE FLOOD SPARGER IN TOP GUIDE STAINLESS STEEL INCORE HOUSING AND GUIDE TUBE STABILIZER CONNECTED TO SHROUD AND SHROUD SUPPORT CONTROL ROD GUIDE TUBE PROVIDES FMCRD BLOW OUT SUPPORT (INTERNAL SUPPORT)
() -
WELDED FEEDWATER SPARGER WITH SECOND THERMAL SLEEVE l
l l
l HEAD VENT j AND SPR AY I
O STEAM DRYER p .-
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- = - = = - - - * +-
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REACTOR ASSEMBLY KEY DESIGN FEATURES
FUEL SYSTEM & THERMAL-HYDRAULIC DESIGN O
THESE ARE AS DESCRIBED IN NRC APPROVED GENERAL ELECTRIC STANDARD APPLICATION FOR REACTOR FUEL, NEDE-24011 (C-LATTICE 8 X 8 FUEL).
IF, AS MAY BE EXPECTED, NEW FUEL DESIGNS ARE APPROVED, IT IS EXPECTED THAT THEY WOULD BE USED IN ABWR.
THE CONTROL ROD IS NEW IN THAT IS HAS NO VELOCITY LIMITER.
THE REASON FOR VELOCITY LIMITER ELIMINATION WILL BE DISCUSSED LATER.
O O
I
FUEL SYSTEM DESIGN O
HANDLE O
c 0 G 0 9 0
- s\ o
'O NEUTRON D 0 ABSORBER O g RODS SHEATHg
,JAQ $ '
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cp -
\
ns 1
COUPLING SOCKET l O
) CONTROL ROD ASSEMBLY
REACTOR INTERNAL MATERIALS O
MATERIAL SPECIFICATIONS - ALMOST ALL REACTOR INTERNAL COMPONENTS ARE AUSTENITIC STAINLESS STEEL TYPE 304L OR 316L PER ASME (FOR CORE SUPPORT STRUCTURES) OR ASTM (FOR OTHER INTERNAL STRUCTURES) MATERIAL SPECIFICATIONS DEPENDING ON PRODUCT FORM.
O SHROUD SUPPORT STRUCTURE WHICH MAKES TRANSITION BETWEEN LOW ALLOY STEEL VESSEL AND STAINLESS STEEL CORE SUPPORT STRUCTURE IS NICKEL-CHROME-IRON (INCONEL 600) AS ARE THE SHROUD HEAD BOLTS.
O CORE PLATE AND TOP GUIDE MOUNTING BOLTS AND PINS ARE PER ASME SA-479 (TYPE 304, 316 OR XM-19)
O O CORE SUPPORT STRUCTURES ARE CONSTRUCTED IN ACCORDANCE WITH ASME CODE SECTION III WITH WELDING PROCEDURES AND WELDERS QUALIFIED PER SECTION IX.
9 O
REACTOR INTERNAL MATERIALS (CONTINUED)
O FABRICATION AND PROCESSING OF AUSTENITIC STAINLESS STEEL -
REGULATORY GUIDES 1.31, 1.37 AND 1.44 ARE COMPLIED WITH IN THAT: )
\
0 BASE MATERIAL IS SOLUTION ANNEALED AS VERIFIED BY TESTING FOR SENSITIZATION 0 COLD FORMING IS CONTROLLED AS VERIFIED BY HARDNESS TEST (WITH THE EXCEPTION OF THE STEAM DRYER VANES, WHICH HAVE ALWAYS BEEN USED AS COLD FORMED)
MATERIALS WITH YIELD STRENGTH IN EXCESS OF 90,000 PSI ARE NOT USED 0 THE DELTA FERRITE CONTENT OF ALL WELDING MATERIALS IS TESTED TO VERIFY AN AVERAGE FERRITE NUMBER OF 8.0 AND
($) S.0 MINIMUM TO PREVENT MICRO-FISSURING (HOT CRACKING)
AND INTERGRANULAR STRESS CORROSION 0 USE OF LOW CARBON MATERIAL AND WELD HEAT INPUT CONTROL l
PREVENTS SENSITIZATION AND HAS BEEN EFFECTIVE IN PREVENTING PROBLEMS. WHERE STRESS IS HIGH OR PROBLEMS HAVE BEEN EXPERIENCED, COMPONENTS ARE SOLUTION HEAT TREATED AFTER WELDING.
0 SPECIAL CARE IS EXERCISED TO AVOID CONTAMINATION OF THE MATERIAL SURFACE BY MATERIALS WHICH COULD CONTRIBUTE TO INTERGRANULAR CRACKING (HALOGENS, SULPHER OR LOW MELTING POINT ELEMENTS)
O
a FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS O
DESIGN BASES 0 SAFETY PROVIDE RAPID INSERTION SCRAM - COMPARABLE TO FAST SCRAM LOCKING PISTON DRIVE POSITION AND INDIVIDUALLY SUPPORT CONTROL ROD PREVENT ROD DROP - SEPARATION DETECTION PREVENT OR LIMIT RATE OF R0D EJECTION IN
() POSTULATED PRESSURE BOUNDARY FAILURE - BRAKE AND 4
CHECK VALVE O POWER GENERATION PROVIDE POSITIONING IN INCREMENTS TO CONTROL POWER AND CORE POWER SHAPING O
FMCRD FEATURES O
DIVERSE MEANS OF ROD INSERTION HYDRAULIC SCRAM AUTOMATIC NUT RUN UP ELECTRIC RUN IN SELECTED CONTROL ROD RUN IN (SCRRI)
NO SCRAM DISCHARGE VOLUME NO COMMON MODE FAILURE MINIMIZED DRIVE CONTAMINATION MINIMIZED ROD DROP CONSEQUENCES
- ROD SEPARATION DETECTION - VIRTUALLY ELIMINATES ROD l
(]) DROP GANGED MOVEMENT REDUCES INDIVIDUAL ROD WORTH IMPROVED PLANT MANUVERABILITY COMPLIMENTS CORE FLOW LOAD FOLLOWING ALLOWS FOR PLANT AUTOMATION REDUCED TIME FOR PLANT START UP ,
NO PISTON SEALS NO MAINTENANCE ON PISTON SEALS PROVEN CORROSION RESISTANT MATERIALS INTERNAL SHOOT OUT SUPPORT
( LESS CONGESTION UNDER VESSEL
I '
DAYONET COUPLINGk ll HOLLOW PISTON
[___ . _ ,
INTERNAL SUPPORT 1
, .b, TOP CUIDE
- j / 1 LABYRINTH SEAL .
BUFFER l
- .C t 3 1N VESSEL HOUSING" ,
- , , FULL-IN SUPPORT l ' '
I f .
MECilANISM OUTER EETE TUBE
!I f 5
l l
l M -
SCRAh! POSITlON SENSING REED BALL SPINDLE _. ' -9 SWITCH GUIDE TUBE SCRAhl POSITION SCRAM LINE INLET g -
SENSING MAGNET l
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f '
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EVALUATION OF ABWR REACTIVITY CONTROL SYSTEM O
BWR LOCKING PISTON DESIGN ABWR FMCRD COMMENTS REACTIVITY CONTROL
& PLANT AUTOMATION RELATIVELY COARSE .7-INCH STEPS FMCRD PROVIDES BUT FULLY ACCEPTABLE PROVIDES CLOSER FINER 6-INCH STEPS. POWER CONTROL & REACTIVITY GANGED CONTROL FOR A MORE UNIFORM CONTROL &
. PLANT AUTOMATION RATE OF POWER FACILITATES l
DIFFICULT CHANGE. GANGED AUTOMATED PLANT OPERATION OPTION OPERATIONS (UP TO 26 ROD GANGS) FOR AUTOMATED PLANT
( MANEUVERING
! SAFETY / ACCIDENT
- ACCUMULATOR OR HYDRAULIC & FMCRD HAS EQUAL REACTOR PRESSURE ELECTRICAL DR SUPERIOR
- HYDRAULIC INSERTION SAFETY INSERTION ONLY (DIVERSE) CHARACTERISTICS I - VELOCITY LIMITER ROD / DRIVE l REQUIRED FOR ROD SEPARATION CAN DROP ACCIDENT BE DETECTED NO VELOCITY LIMITER REQUIRED.
l
' NO SCRAM DISCHARGE '
VOLUME IN REACTOR BUILDING.
(]) POSITIVE ROD COUPLING.
i EVALUATION OF ABWR REACTIVITY CONTROL SYSTEM O
BWR LOCKING PISTON DESIGN ABWR FMCRD COMMENTS HYDRAULIC DESIGN INDIVIDUAL SCRAM TWO DRIVES PER ELIMINATION OF ACCUMULATORS FOR ACCUMULATOR APPROXIMATELY 100 EACH DRIVE HCU'S SIMPLIFIES THE DESIGN AND REDUCES SURVEILLANCE /
MAINTENANCE REQUIREMENTS TECHNOLOGY BASE
() ESTABLISHED BWR FMCRD TECHNOLOGY SEE FOLLOWING TECHNOLOGY HAS BEEN FIRMLY
SUMMARY
ESTABLISHED BY EXTENSIVE EUROPEAN EXPERIENCE AND ABWR T&D PROGRAMS FMCRD TECHNOLOGY IS WELL ESTABLISHED AND OFFERS MANY ADVANTAGES O
=. . -. .
O FINE MOTION CONTROL ROD DRIVE OPERATING EXPERIENCE AND T&D BASE OPERATING EXPERIENCE e Nearly 2700 Drives in Service in European BWRs 9 Approaching 15,000 Drive-Years of Experience 9 No Forced Outages (During Commercial Operation)
Due to Mechanical Malfunction of FMCRD 9 For the KWU Design Routine Inspections Have Confirmed:
- No Parts Have to be Replaced Because of Wear
- No Observed Performance Degradation (Scram
'nmes, Settling 'Umes, Etc.)
- Once Per 40 Year Inspection is Fully Adequate T&D BASE l 9 Extensive European Testing Over the Years by NSSS Suppliers e Separate ABWR Drive Development Testing PROVEN TECHNOLOGY Q .
i 1
. l CRD TESTING AND VERIFICATION
()
DEVELOPMENT TESTS O PROTOTYPE FMCRD BASED ON EUROPEAN DRIVE DESIGN USED IN :
l OPERATING PLANT 600 SCRAMS 67,000 MOTOR DRIVEN CYCLES 0 SUBSEQUENT PROTOTYPE TO DEMONSTRATE FMCRD PERFORMANCE UNDER ACTUAL PLANT CONDITIONS S00 SCRAMS 63,000 STEP CYCLES CURRENTLY INSTALLED AT LASALLE UNIT 1
)
0 DESIGN PROTOTYPE DESIGN 1,000 SCRAMS 150,000 STEP CYCLES SEISMIC SCRAMABILITY TESTING O RESULTS N0* ABNORMAL DISTORTION OR DEFORMATION NEGLIGIBLE WEAR
- SCRAM TIME REQUIREMENTS SATISFIED
()
O I l
ABWR SSAR CHAPTER 4 REACTOR DESIGN PRESENTED TO ACRS SUBCOMMITTEE ON THE O ADVANCED BOILING WATER REACTOR l
l JUNE 1, 1988 WASHINGTON, D.C.
(
e L
O GE NUCLEAR ENERGY
ACRS SUBCOM'MITTEE ON THE ADVANCED BOILING WATER REACTOR l
(]) JUNE 1, 1988 4
t' CHAPTER 4 -
REACTOR OVERVIEW CONSISTENT WITH THE SRP CHAPTER 4 ADDRESSES:
O REACTOR ASSEMBLY (RPV & INTERNALS)
FUEL SYSTEM DESIGN REACTOR THERMAL-HYDRAULIC DESIGN REACTOR MATERIALS REACTIVITY CONTROL SYSTEMS O
6 REACTOR VESSEL AND INTERNALS O DESIGN OBJECTIVES 0 DEFINE A REACTOR SYSTEM WHICH FULLY CONTRIBUTES TO ALL MAJOR ABWR OBJECTIVES
- DESIGN SIMPLIFICATION
- ADDITIONAL MARGINS
- REDUCED MAINTENANCE COSTS
- LOWER OCCUPATIONAL EXPOSURE O - CAPACITY FACTOR / AVAILABILITY IMPROVEMENTS
- ENHANCED REACTOR SAFETY 0 ADDRESS ALL BWR MATERIALS CONCERNS 0 USE ESTABLISHED TECHNOLOGY OR TEST BEFORE USE
()
s _-__ 2 -m a. 2 sn. m - - - ~ 1 l
REACTOR KEY DESIGN FEATURES
() 0 REACTOR PRESSURE VESSEL ,
- REDUCED RADIUS CLOSURE HEAD AND FLANGES RIP PENETRATIONS MACHINED IN BOTTOM HEAD KNUCKLE FORGING STEAM FLOW RESTRICTOR IN N0ZZLE-ALSO USED FOR FLOW MEASUREMENT RPV VENT AND HEAD SPRAY N0ZZLE IN TOP HEAD 0 REACTOR INTERNALS BWR/6 TYPE STEAM DRYER AND STEAM SEPARATORS BORON CARBIDE CONTROL ROD WITHOUT VELOCITY
() LIMITER (1 FOOT SHORTER)
BWR/6 TYPE CORE STRUCTURE WITH TOP GUIDE MACHINED FRDM THICK PLATE AND CORE PLATE STIFFENERS IN TWO DIRECTIONS ,
SEPARTE LOW PRESSURE FLOOD SPARGER IN VESSEL AND HIGH PRESSURE CORE FLOOD SPARGER IN TOP GUIDE STAINLESS STEEL INCORE HOUSING AND GUIDE TUBE :
STABILIZER CONNECTED TO SHROUD AND SHROUD SUPPORT CONTROL ROD GUIDE TUBE PROVIDES FMCRD BLOW OUT SUPPORT (INTERNAL SUPPORT) 0 -
WELDED FEEDWATER SPARGER WITH SECOND THERMAL SLEEVE
HE AD VENT
i t i l ll ,
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REACTOR ASSEMBLY KEY DESIGN FEATURES
FUEL SYSTEM & THERMAL-HYDRAULIC DESIGN O
THESE ARE AS DESCRIBED IN NRC APPROVED GENERAL ELECTRIC STANDARD APPLICATION FOR REACTOR FUEL, NEDE-24011 (C-LATTICE 8 X 8 FUEL).
IF, AS MAY BE EXPECTED, NEW FUEL DESIGNS ARE APPROVED, IT IS EXPECTED THAT THEY WOULD BE USED IN ABWR.
THE CONTROL ROD IS NEW IN THAT IS HAS NO VELOCITY LIMITER.
THE REASON FOR VELOCITf LIMITER ELIMINATION WILL BE DISCUSSED LATER.
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FUEL SYSTEM DESIGN O
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REACTOR INTERNAL MATERIALS O
1 MATERIAL SPECIFICATIONS - ALMOST ALL REACTOR INTERNAL l COMPONENTS ARE AUSTENITIC STAINLESS STEEL TYPE 304L DR 316L PER ASME (FOR CORE SUPPORT STRUCTURES) OR ASTM.(FOR OTHER l INTERNAL STRUCTURES) MATERIAL SPECIFICATIONS DEPENDING ON PRODUCT FORM.
O SHROUD SUPPORT STRUCTURE WHICH MAKES TRANSITION BETWEEN LOW ALLOY STEEL VESSEL AND STAINLESS STEEL CORE SUPPORT STRUCTURE IS NICKEL-CHROME-IRON (INCONEL 600) AS ARE 1
THE SHROUD HEAD BOLTS.
t 0 CORE PLATE AND TOP GUIDE MOUNTING BOLTS AND PINS ARE PER ASME SA-479 (TYPE 304, 316 OR XM-19)
O CORE SUPPORT STRUCTURES ARE CONSTRUCTED IN ACCORDANCE l
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WITH ASME CODE SECTION III WITH WELDING PROCEDURES AND WELDERS QUALIFIED PER SECTION IX.
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l f REACTOR INTERNAL MATERIALS (CONTINUED)
O FABRICATION AND PROCESSINC t OF AUSTENITIC STAINLESS STch REGULATORY GUIDES 1.31. 1.37 AND 1.44 ARE COMPLIED WITH IN ,
THAT:
0 BASE MATERIAL IS SOLUTION ANNEALED AS VERIFIED BY TESTING FOR SENSITIZATION O COLD FORMING IS CONTROLLED AS VERIFIED BY HARDNESS TEST (WITH THE EXCEPTION OF THE STEAM DRYER VANES, WHICH HAVE ALWAYS BEEN USED AS COLD FORMED)
MATERIALS WITH YIELD STRENGTH IN EXCESS OF 90,000 PSI ARE NOT USED 0 THE Dit.TA FERRITE CONTENT OF ALL WELDING MATERIALS IS TESTED TO VERIFY AN AVERAGE FERRITE NUMBER OF 8.0 AND O 5.0 MINIMUM TO PREVENT MICRO-FISSURING (HOT CRACKING)
AND INTERGRANULAR STRESS CORROSION O USE OF LOW CARBON MATERIAL AND WELD HEAT INPUT CONTROL PREVENTS SENSITIZATION AND HAS BEEN EFFECTIVE IN PREVENTING PROBLEMS. WHERE STRESS IS HIGH OR PROBLEMS HAVE BEEN EXPERIENCED, COMPONENTS ARE SOLUTION HEAT TREATED AFTER WELDING. !
O SPECIAL CARE IS EXERCISED TO AVOID CONTAMINATION OF THE MATERIAL SURFACE BY NATERIALS WHICH COULD CONTRIBUTE TO INTERGRANULAR CRACKING (HALOGENS, SULPHER OR LOW MELTING POINT ELEMENTS) i 0
FUNCTIONAL DESIGN OF REACTIVITY CONTROL SYSTEMS O
DESIGN BASES 0 SAFETY PROVIDE RAPID INSER110M 3 CRAM - COMPARABLE TO FAST SCRAM LOCKING PIST0ff CRIVE POSITION AND INDIVIDUALLY SUPPORT CONTROL ROD PREVENT ROD DROP - SEPARATION DETECTION PREVENT OR LIMIT RATE OF ROD EJECTION IN
([) POSTULATED PRESSURE BOUNDARY FAILURE - BRAKE AND CHECK VALVE t
0 POWER GENERATION PROVIDE POSITIONING IN INCREMENTS TO CONTROL POWER AND CORE POWER SHAPING 4
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1 FMCR0 FEATURES O
DIVERSE MEANS OF ROD INSERTION HYDRAULIC SCRAM AUTOMATIC NUT RUN UP ELECTRIC RUN IN SELECTED CONTROL ROD RUN IN (SCRRI)
NO SCRAM DISCHARGE VOLUME NO COMMON MODE FAILURE MINIMIZED DRIVE CONTAMINATION MINIMIZED ROD DROP CONSEQUENCES ROD SEPARATION DETECTION - VIRTUALLY ELIMINATES ROD DRDP
() GANGED MOVEMENT REDllCES INDIVIDUAL ROD WORTH IMPROVED FLANT MANOVERABILITY COMPLIMENTS CORE FLOW LOAD FOLLOWING i ALLOWS FOR PLANT AUTOMATION
{ REDUCED TIME FOR PLANT START UP l
NO PISTON SEALS NO MAINTENANCE ON PISTON SEALS 1
PROVEN CORROSION RESISTANT MATERIALS INTERNAL SHOOT OUT SUPPORT
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l EVALUATION OF ABWR REACTIVITY CONTROL SYSTEM l
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() BWR LOCKING PISTON DESIGH ABWR FMCRD COMMENTS REACTIVITY CONTROL
& PLANT AUTOMATION 1 i
I RELATIVELY COARSE .7-INCH STEPS FMCRD PROVIDES BUT FULLY ACCEPTABLE PROVIDES CLOSER FINER 6-INCH STEPS. POWER CONTROL & REACTIVITY GANGED CONTROL FOR A MORE UNIFORM CONTROL &
PLANT AUTOMATION RATE OF POWER FACILITATES DIFFICULT CHANGE. GANGED AUTOMATED PLANT OPERATION OPTION OPERATIONS (UP TO 26 ROD GANGS) FOR AUTOMATED PLANT
(]) MANEUVERING SAFETY / ACCIDENT
- ACCUMULATOR OR HYDRAULIC & FMCRD HAS EQUAL REACTOR PRESSURE ELECTRICAL OR SUPERIOR
- HYDRAULIC INSERTION SAFETY INSERTION ONLY (DIVERSE) CHARACTERISTICS
- VELOCITY LIMITER ROD / DRIVE REQUIRED FOR R0D SEPARATION CAN ,
DROP ACCIDENT BE DETECTED NO VELOCITY LIMITER REQUIRED.
NO SCRAM DISCHARGE VOLUME IN REACTOR BUILDING.
(I POSITIVE ROD COUPLING.
l EVALUATION OF ABWR REACTIVITY CONTROL SYSTEM O
BWR LOCKING PISTON DESIGN ABWR FMCRD COMMENTS HYDRAULIC DESIGN INDIVIDUAL SCRAM TWO DRIVES PER ELIMINATION OF ACCUMULATORS FOR ACCUMULATOR APPROXIMATELY 100 EACH DRIVE HCU'S SIMPLIFIES THE DESIGN AND REDUCES SURVEILLANCE /
MAINTENANCE REQUIREMENTS TECHNOLOGY BASE l
(]} ESTABLISHED BWR FMCRD TECHNOLOGY SEE FOLLOWING TECHNOLOGY HAS BEEN FIRMLY
SUMMARY
l ESTABLISHED BY EXTENSIVE EUROPEAN EXPERIENCE AND
( ABWR T&D PROGRAMS l
FMCRD TECHNOLOGY IS WELL ESTABLISHED AND OFFERS MANY ADVANTAGES CI
FINE MOTION CONTROL ROD DRIVE O
i OPERATING EXPERIENCE AND T&D BASE ,
OPERATING EXPERIENCE e Nearly 2700 Drives in Service in European BWRs e Approaching 15,000 Drive-Years of Experience e No Forced Outages (During Commercial Oper: tion)
Due to Mechanical Malfunction of FMCRD 9 For the KWU Design Routine Inspections Have Confirmed:
- No Parts Have to be Replaced Because of Wear
- No Observed Performance Degradation (Scram
'I12nes, Settling 'Ilmes, Etc.)
- Once Per 40 Year Inspection is Fully Adequate O T&D BASE 9 Extensive European Testing Over the Years by NSSS Suppliers e Separate ABWR Drive Development Testing -
4 4
PROVEN TECHNOLOGY
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CRD TESTING AND VERIFICATION CE)
DEVELOPMENT TESTS 0 PROTOTYPE FMCRD BASED ON EUROPEAN DRIVE DESIGN USED IN OPERATING PLANT 600 SCRAMS 67,000 MOTOR DRIVEN CYCLES 0 SUBSEQUENT PROTOTYPE TO DEMONSTRATE FMCRD PERFORMANCE UNDER ACTUAL PLANT CONDITIONS 500 SCRAMS 63,000 STEP CYCLES CURRENTLY INSTALLED AT LASALLE UNIT 1 0
0 DESIGN PROTOTYPE DESIGN 1,000 SCRAMS 150,000 STEP CYCLES SEISMIC SCRAMABILITY TESTING O RESULTS
- NO ABNORMAL DISTORTION OR DEFORMATION
- NEGLIGIBLE WEAR SCRAM TIME REQUIREMENTS SATISFIED O'
._ -